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REQUIREMENTS FOR THE CONTENT OF THE SAFETY ANALYSIS REPORT

FOR NUCLEAR POWER PLANTS WITH FAST REACTORS

NP-018-05

ABBREVIATIONS

ASB - Automatic Switch-over to BackupEP - Emergency ProtectionNFME - Neutron Flux Monitoring EquipmentNPP - Nuclear Power PlantARSMS - Automatic Radiation Situation Monitoring SystemSFP - Spent Fuel PoolBN - Fast Neutron Reactor PlantSFAD - Spent Fuel Assembly DrumMCR - Main Control RoomBRU-A - Steam Dump Valve to AtmosphereBRU-K - Steam Dump Valve to CondenserPSA - Probabilistic Safety AssessmentES - Explosive SubstancesRI – Reactor InternalsICM - In-Core MonitoringAHE - Air Heat ExchangerITPS (TPS) - In-Plant Transportation Package SetASW - Air Shock WaveWCR - Water Chemistry RegimeCD - Civil DefenceRCP - Reactor Coolant PumpUNARSMS - Unified National Automated Radiation Situation Monitoring SystemLRW - Liquid Radioactive WastesMBA - Material Balance AreaPSS - Protective Safety Systems- - ActuatorIE - Initiating EventPRZ - PressurizerSC - Short Circuit- - InstrumentationI&C - Instrumentation, Control and Automation SystemsLSS - Localizing Safety SystemsMSW - Maximum Surge WaveEMERCOM of Russia - Ministry of the Russian Federation for Civil Defense, Emergency Situations and Disaster ReliefMCL - Minimum Controlled LevelSSE - Safe Shutdown EarthquakeMSK-64 - Earthquake ScaleRD - Regulatory DocumentS&RW - Scientific and Research WorksSVRE - Sodium Void Reactivity EffectRSS - Radiation Safety StandardsGH - Geological HazardR&DW - Research and Development WorksNPP SAR - Nuclear Power Plant Safety Analysis Report

FSAR - Final Safety Analysis ReportGSP - General Safety Provisions for Nuclear Power PlantsSSS - Supporting Safety SystemsOST - Industry-Specific StandardSNF - Spent (Irradiated) Nuclear FuelPEP - Preventive Emergency ProtectionNPP RF NSR - Nuclear Safety Rules for Reactor Facitities of Nuclear Power PlantsSG - Steam GeneratorDBE - Design-Basis EarthquakePADS - Potential Accidental Detonation SourcesPIE - Postulated Initiating EventSV - Safety ValveCW - Commissioning WorksNPP QAP - Nuclear Power Plant Quality Assurance ProgramPSAR - Preliminary Safety Analysis ReportSPM - Scheduled Preventive MaintenanceNPU Rules - Rules for Arrangement and Safe Operation of the Equipment and Pipelines for Nuclear Power UnitsST - Software ToolsIHE - Intermediate Heat ExchangerAE - Absorber ElementRW - Radioactive WastesRS - Radiation SafetyRSb - Radioactive SubstancesGARW - Gas and Aerosol Radioactive WastesSDGS - Standby Diesel Generator StationRB - Reactor Building CR - Control RodCPS CR - Control Rod of the Control and Protection SystemECR - Emergency Control RoomTG - Technical GuidesRP - Reactor PlantSG EPS - Steam Generator Emergency Protection SystemERCS - Emergency Reactor Cooldown SystemEAS - Emergency Alarm SystemEPSS - Emergency Power Supply SystemsSS - Safety SystemsSRS - Safety-Related SystemsICIS - In-Core Instrumentation SystemLCS - Leak-Tight Containment SystemSPA - Sanitary-Protective AreaSPS - Seismic Protection SystemNPP SR - Sanitary Rules for Design and Operation of Nuclear Power PlantsCS - Control SystemCPS - Control and Protection SystemPPS - Physical Protection SystemSSCR - Self-Sustaining Chain ReactionFA - Fuel AssemblyFE - Fuel ElementTC - Transport ContainerSCC - Short-Circuit CurrentTDFP - Turbine-Drived Feedwater PumpSRW - Solid Radioactive WastesTP - Turbine PlantAM - Accident ManagementPSE - Pressure Safety ElementCSS - Control Safety SystemsSNFS - Spent Nuclear Fuel StorageFFS - Fresh Fuel Storage- - Operating Organization

NM - Nuclear MaterialsNF - Nuclear Fuel

I. GENERAL REQUIREMENTS

1. Purpose and application scope of the regulatory document "Requirements for the content of the safety analysis report for nuclear power plants with fast reactors"

1.1. This RD shall be applicable to any legal entities and natural persons performing any activities for siting, construction and operation of NPPs with fast reactors using liquid-metal sodium as the coolant.

1.2. This RD shall establish the requirements for:

- the purpose and application scope of the NPP SAR;

- the procedure for the report preparation;

- the report issuance and maintenance;

- standard description of individual NPP systems in the report.

2. Purpose and application scope of the report

2.1. The NPP SAR is the document substantiating safety assurance for the NPP power unit.

2.2. The NPP SAR shall contain sufficient information to understand the NPP power unit design, safety concept the design is based on, the NPP QAP and the basic operation principles proposed by the Applicant.

Based on the information contained in the NPP SAR the state regulatory authority for safe atomic energy use shall be able to assess adequacy of safety analysis in the course of siting, construction, commissioning, operation and decommissioning of the NPP power unit at the particular site in order to prevent any exceedance of the established exposure doses for the workers and the public and any norms of radioactive substance releases, discharges and content in the environment under normal operation conditions and in case of any design basis accidents as well as to provide the possibility to limit radiation exposure in case of any beyond design basis accidents.

2.3. An individual NPP SAR shall be developed for each power unit of multi-unit NPPs.

2.4. Requirements for the content of the NPP SAR are developed for NPPs using liquid-metal sodium as the coolant. Provisions of the report may be also applicable to NPP power units with fast reactors using other types of coolant. Peculiarities of these NPPs and their difference from NPP power units with reactors using liquid-metal sodium as the coolant shall be taken into consideration in order to apply the requirements.

3. Procedure for the report preparation

3.1. Preparation and formation of the NPP SAR shall be performed at all stages of the NPP lifecycle.

The data used in the NPP SAR shall correspond to the status of the power plant in accordance with the design documentation and its actual condition.

3.2. The information contained in the PSAR shall be based on the materials of the NPP design, the RP and equipment designs, the results of surveys, S&RW and R&DW.

The information contained in the NPP SAR shall reflect the actual condition of the NPP (the NPP power unit) according to the results of construction, installation, precommissioning works and inspections, physical and power start-up and testing of the power unit in its entirety.

3.3. Subsequent to completion of all commissioning works for the NPP (the NPP power unit) the NPP SAR shall be adjusted.

3.4. All modifications of the design shall be reflected in the report and included into the report.

4. Requirements for the report content and presentation

Content and presentation of the NPP SAR shall comply with the requirements stated in this regulatory document.

4.1. Requirements for the content of the report

4.1.1. The content of the NPP SAR shall where practicable be such that the state regulatory authority for safe atomic energy use would not have to review any additional design, engineering and operational documents for the purpose of safety assessment.

4.1.2. The structure of the PSAR and FSAR shall be uniform.

4.1.3. The information shall be presented in a clear and understandable way. It shall be consistent in comparison of different subsections. The information on compliance with the requirements shall not have declarative character. Documented evidence of such compliance shall be presented.

In case the information is based on any studies or documents the reference to these studies or documents shall be given with indication of the document type, the authors or organization, the year of performance or issuance, the archive or identification number.

It is recommended to follow the description structure specified in Section 5 for presentation of the information on any systems.

4.1.4. Repetitions of the information shall be avoided. Developers of the NPP SAR shall make the decision on the necessity of any repetitions based on convenience of the information perception and assessment as well as on the quantity of information. It is recommended to give references to the relevant sections in order to avoid excessive repetitions. In case the description of any individual safety systems is given in the NPP SAR sections intended for presentation of normal operation systems these descriptions may be repeated completely in Section 12.

4.1.5. The information on any performed calculations and design analyses shall confirm sufficiency and completeness of the performed calculations, consideration of all factors affecting the result and shall also contain the data sufficient to perform the expert calculation (where necessary) (diagrams, assumptions, input data, results, their interpretation and conclusions).

4.1.6. All software tools shall be briefly described within the scope sufficient to understand and assess their applicability; their names and validation data shall be specified.

4.1.7. Lists of bibliographical references shall be given in each section.

4.2. Requirements for the presentation of the report

4.2.1. Presentation of the report shall be uniform for all stages and all sections. The Applicant shall compile the NPP SAR by separate sections and sub-sections in folders.

The NPP name, the full name of the NPP SAR and the relevant section shall be indicated on each folder.

4.2.2. The first folder shall contain:

- the contents of the entire NPP SAR;

- introduction, Section 1;

- general information (abstract, abbreviations).

The contents of the entire NPP SAR shall be presented in each folder.

The NPP SAR shall be arranged in compliance with the requirements for presentation of text documents.

4.2.3. Any modifications to the NPP SAR text shall be introduced by correction or replacement of individual pages or subsections. The revision history sheet shall be placed in the end of each section.

4.2.4. The standard system description structure for the NPP SAR is given in Appendix 1.

II. SAFETY ANALYSIS REPORT FOR NPPs WITH FAST REACTORS

REQUIREMENTS FOR THE INTRODUCTION

General information on the NPP design, developers of the power plant design and the NPP SAR, stages of the entire design development as well as general characteristics of the NPP SAR shall be provided.

1. Design development basis

Brief information on the decisions used as the basis for the planned NPP construction and design development shall be provided.

2. General characteristics of the NPP

The general characteristics of the NPP particularly its purpose, the list of electric power and heat consumers, brief characteristics of the site and its peculiarities, the planned NPP capacity, the number of power units, the expected commissioning schedule for the designed power unit, the NPP operation modes, the main electric connection diagram, etc. shall be presented.

The general characteristics of the power unit including its thermal and power capacity, the auxiliary heat extraction and power consumption values, the configuration of the power unit (the number of circuits, the type of coolant and the fluid), brief description of the circuit process systems and their main process equipment, the principal heat balance diagram and description of the heat removal principle, the circuit parameters (pressure, temperature, media flow rates), characteristics of the unit power supply system, layout of the basic structures and equipment (the RB, the turbine hall), the list of safety systems and their classification into protection, localizing, support and control ones, description of the primary circuit equipment, brief description of the nuclear core (the number of operating, breeder and protection fuel assemblies, the CPS number and type, NF burnup) and the fresh NF used

shall be provided.

3. Report development stage

The purpose and summary of the NPP SAR as well as the licensing process stage the presented NPP SAR is developed for shall be indicated.

4. Information on the operating organization and contractors

Information on the operating organization and the list of the basic organizations engaged in design, construction, manufacturing and installation of the main safety-related systems and equipment of the plant shall be provided.

The basic organizations rendering any services to the operating organization in the course of the power unit construction and operation shall be specified. References to the relevant administrative and qualification documentation shall be given, and allocation of functions and responsibilities for these organizations shall be provided.

5. Information on the report developers

Information on the operating organization submitting the NPP SAR to the state regulatory authority for safe atomic energy use and on the developers of individual independent sections of the NPP SAR, particularly information on their working experience, availability of licenses for the activities in the area of atomic energy use granted by the state regulatory authority for safe atomic energy use shall be specified.

6. Information on S&RW and R&DW

Brief information on S&RW and R&DW performed or planned to be performed for substantiation of the equipment designs, safety and the main design solutions shall be presented.

7. Characteristics of the report

Completeness of the presented information and its compliance with the requirements of this RD shall be characterized.

In case the design development is at the initial stage and thus the presented information does not comply with the requirements of this RD to the full extent this circumstance shall be indicated in this section.

1. REQUIREMENTS FOR THE SECTION "GENERAL DESCRIPTION OF THE NPP POWER UNIT"

Information on the contents of all NPP SAR sections shall be presented.

Peculiarity of the information shall consist in the possibility to use it independently regardless of any other NPP SAR sections, particularly for familiarization of the local authorities, public organizations and the population with the concept and the basic technical solutions for the NPP power unit safety assurance. Information shall be presented in a simple and understandable manner. However it should be not mechanically abridged information from other sections, but independent statement with tables, diagrams and drawings.

1.1. Construction conditions

Brief information on the NPP site and its location area shall be provided.

- Climatic conditions.

- Characteristics of the atmosphere.

- Ambient air temperatures: average monthly for several years, extreme per a year, the maximum of the average monthly, average ten-day and one-time ones.

- Geological, hydrogeological and seismotectonic characteristics.

- Seismicity of the NPP site area, the boundaries of the solid block without any seismic deformations particularly in case of SSE.

- Characteristics of the extreme natural phenomena (whirlwinds, hurricanes, tornadoes, dust storms, icing, flooding, etc.).

- Characteristics of soils to the depth of at least 100 m with indication of the distribution of compressible (clayey, sandy) and non-compressible (rocky and semi-rocky) soils.

- Depth of the first aquifer from the surface and its connections to the surface waters.

- Data on the population density within the radius of 25 km from the NPP including engaged and operating personnel of the NPP.

- Data on the SPA and the number of populated localities subject to relocation prior to the NPP commissioning.

1.2. Location plan

Brief description of the NPP site location area including brief characteristics and location of the facilities, water conduits of the pump stations, water reservoirs, irrigation channels, dams of the hydro power plants, air fields, motorways and railways with indication of their position in relation to the SPA and the supervised area shall be provided.

The site terrain and inclinations towards any water bodies shall be characterized. Brief information on land use shall be presented.

Directions of the high-voltage power transmission lines of the NPP, any access motorways and railways and the expected location of the housing settlement (for the NPP workers) shall be indicated.

Any facilities of particular hazard with regard to explosions, fires and releases of toxic substances into the environment shall be specified. The location plan on a scale of 1:25 000 shall be provided.

1.3. Description of the NPP power unit flow diagram

The flow diagram of the power unit with indication of the following shall be provided:

1. Primary circuit

1.1. Reactor

1.2. Pressure compensation system with safety valves

1.3. Primary circuit coolant monitoring and purification system

1.4. Gas systems of the primary circuit

1.5. Pipelines

2. Secondary circuit

2.1. Pipelines

2.2. SG

2.3. Secondary circuit RCP

2.4. Buffer tank

2.5. Secondary circuit coolant receipt, accumulation and purification system

2.6. Gas systems of the secondary circuit

3. ERCS

4. SFP with the cooling system

5. Steam pipelines

6. Steam turbine plant

7. Feed train

8. System for cooldown and residual heat removal via the third circuit

9. Service water supply system for the normal operation systems and the safety systems

1.4. Basic technical characteristics of the NPP power unit

The basic technical characteristics of the power unit shall be specified, including:

1. Power and thermal capacity

2. Heating capacity

3. Capacity factor

4. Auxiliary power consumption

5. NF loading

6. RP service life

7. Main parameters of the circuit coolant

8. Other parameters required to understand the basic characteristics of the power unit

1.5. Grid characteristics

The principal scheme of the grid where the NPP power unit is to be operated as well as the following information on the grid shall be provided:

1. Voltage in the grid networks

2. State of the grid as of the NPP power unit start-up with indication of the type and capacity of power plants in the grid

3. General levels of power consumption and maximal loads in the grid (daily, weekly, seasonal and annual), capacity margins in relation to the maximal loads

4. Operation modes of the grid automation and protection devices affecting the NPP power unit operation mode

5. Operation of the NPP power unit in case of any malfunctions in the grid resulting in the power unit load shedding down to the auxiliary

In case of any expected malfunctions it is necessary to determine the time for restoration of the auxiliary NPP unit power supply from the external source.

1.6. Power unit operation modes and events considered in the power unit design

Information on all NPP power unit operation modes and design basis accidents including the events caused by external impacts with the frequency of once per 100 years as well as the impact of SSE, ASW and aircraft crash shall be provided.

1.7. Safety assurance concept

1.7.1. Basic safety assurance principles and criteria

The following shall be provided:

1. The list of effective federal laws and RDs in the area of safety Only basic safety documents for the NPP may be provided, and the complete list may be presented in the appendix to this section.

2. Quantitative values of the safety criteria underlying the power unit design

3. Description of safety assurance through consistent implementation of the defense-in-depth principle based on the system of physical barriers in the way of ionizing radiation and radioactive substance propagation into the environment and the multi-level system of technical and administrative measures for protection of the barriers and maintenance of their efficiency as well as for protection of the public.

4. Information on application of the internal self-protection principle in the design (the means to implement it)

5. Information on the solutions provided in the NPP power unit design in order to ensure the relevant safety level

6. The list of safety systems and the main functions performed by these systems

7. Confirmation of compliance with the basic SS arrangement principles, particularly:

- passivity;

- single failure;

- multiple channels;

- physical separation;

- diversity;

- in-service inspection.

8. Confirmation of the SS robustness against common cause failures (fires, blackout, external natural and human-induced impacts, etc.)

9. Confirmation of the SS robustness against human errors

10. Information on performance of the SS functions under the impact of earthquake, ASW and aircraft crash on the power unit

11. Information on beyond design basis accidents (the list of considered beyond design basis accidents; any measures aimed to mitigate their consequences; severe accident management measures)

12. Information on the experience in design, construction, installation, operation and testing confirming adequacy of the technical and administrative solutions adopted in order to ensure the power unit safety

1.7.2. Nuclear safety assurance

The nuclear safety objectives shall be formulated, and the systems used to achieve these objectives shall be indicated, particularly:

- SSCR control in the nuclear core of the reactor (it should be demonstrated how far the nuclear safety is based on application of the internal self-protection properties of the reactor) The data on reactivity balance for all possible operational states, emergency situations and design basis accidents shall be provided. It is recommended to present the data in tabular format. The structure of the provided technical reactivity control means, functions of individual systems and subsystems and their reliability shall be described. The means to ensure compliance with the requirements of the NPP RF NSR shall be demonstrated. Information on the efficiency, reliability and response rate of the reactor EP shall be specified.

- Heat removal from the reactor nuclear core.

- Prevention of local criticality in the course of NF reloading, transportation and storage (brief information on prevention of local criticality in the course of the above-mentioned works)

1.7.3. Radiation safety assurance

Information on the engineering features and administrative measures aimed to ensure protection of the personnel, the public and the environment against radiation exposure shall be specified. It should be demonstrated that application of the proposed protection means and implementation of the protective measures are justified by practice and do not result in exceedance of the established dose limit, prevent any unsound irradiation, and the existing radiation exposure is maintained at the lowest reasonably achievable level with due regard for the economic and social factors.

1.7.4. Fire safety assurance

General criteria of safety assurance in case of any fire at the power unit shall be specified. The following shall be considered:

- a fire as an IE or consequence of any IE with due regard for the single failure principle (in the fire extinguishing systems);

- a fire as an IE coinciding with any other IE with probability assessment for such coincidence.

The fire safety assurance concept shall be formulated, and its criteria shall be specified.

Multiple barriers, the optimal balance between the passive and active protection and adequacy of the fire extinguishing systems shall be substantiated.

Qualitative assessment of fire consequences with due regard for any potential failures of the fire extinguishing units shall be performed (quantitative assessment of the fire impact on safety shall be performed within the PSA. Results of this assessment may be presented in Subsection 1.8.3).

Systematic approach to fire safety assurance in the course of fire safety measures shall be demonstrated.

Fire hazards in various areas of the power unit shall be analyzed. The zoning principle for the buildings (division into fire zones and compartments) shall be described. The buildings shall be classified in accordance with their fire resistance depending on the fire hazard category of the building and its importance for safety assurance.

Impossibility to lose more than one SS train due to a fire shall be substantiated.

Compliance with the RD requirements for fire safety shall be demonstrated.

The following data on the active fire extinguishing systems shall be provided:

- the arrangement principle for these systems;

- their reliability level;

- operability analysis for these systems in case of any single failures;

- analysis of extreme impacts on the fire detection and extinguishing means.

Capacity and selection of the filters for purification of the combustion products shall be substantiated for the emergency fire ventilation systems.

Consequences of spurious actuation of the fire extinguishing units (impact on the safety-related equipment) shall be analyzed.

The design quantity of simultaneous fires at the site shall be determined, and adequacy of the fire extinguishing systems and means for the NPP fire safety assurance shall be substantiated.

It should be demonstrated that external fires at the production site will not affect the workers, the civil structures, any buildings located near the fire and safety-related equipment which is to be operable within this period.

1.7.5. Protection against natural and human-induced impacts

The following information shall be provided:

- for the SRS structures, assemblies and equipment - the list of extreme impacts with the

frequency of 10*(-2) 1/year (winds, hurricanes, tornadoes, whirlwinds, extreme temperatures, floods, icing, etc.) with indication of the impact magnitude as well as parameters of impact caused by aircraft crash, missiles and ASW;

- measures for protection against external impacts;

- characteristics of earthquakes and their parameters taken into account for design of buildings and structures referred to the first and the second category, information on the SPS;

- hazards of the industrial, transportation and military facilities located in the vicinity of the NPP. Any potential sources of explosion accidents at these facilities and ASW impact parameters;

- regulatory basis for calculations of the protection against external impacts, methods and calculation programs for assessment of external impacts.

1.7.6. Action plans for the personnel and the public protection in case of any accident

The basic provisions of the action plans for the personnel and the public protection in case of any radiation accident at the NPP power unit shall be presented.

The procedure for the public notification shall be specified, and the organizational measures in case of any accident including coordination of actions between the NPP workers and the facility-based, federal and territorial forces of the Russian EMERCOM shall be presented.

1.8. Safety analysis results

1.8.1. Reliability of the equipment and other components

Information on reliable performance of the NPP power unit safety functions (shutdown and maintenance of the reactor in the shutdown state, emergency cooling of the reactor, confinement of radioactive substances) shall be provided.

1.8.2. Deterministic safety analysis

Brief information on the performed safety analyses described in detail in Section 15 shall be provided.

The information shall be presented for all groups of the considered accidents.

General assessment of the obtained NPP safety analysis results shall be given.

1.8.3. Probabilistic safety analysis

Information on the performed PSA (Level 1 PSA for construction of the NPP power unit and Level 1 PSA adjusted subsequent to the commissioning results - for operation) including the PSA results with assessment of their compliance with the GSP requirements shall be provided.

1.9. Basic design solutions

1.9.1. Reactor, primary circuit and associated systems

The following brief information shall be provided:

- general description of the reactor, the primary circuit and the associated systems including description of the reactor installation in the cavity, biological and radiation protection, purpose of individual systems and components;

- classification of the components of the reactor, the primary circuit and the associated systems (reference shall be given to the section containing the classification);

- basic operational characteristics of the systems and components;

- principles and criteria underlying the design of the reactor, the primary circuit and the associated systems.

Description shall be accompanied with the process flow diagrams and drawings (the reactor installation in the cavity, the assembled reactor, the section across the core, the main components of the core, the reactor pressure vessel, the RCP, the SG, the pressurizer, kinematic diagram of the CPS drive).

1.9.2. Secondary circuit


- general description of the secondary circuit, the SG and the associated systems;

- classification of the components of the secondary circuit and the associated systems;

- basic operational characteristics of the secondary circuit systems and components;

- principles and criteria underlying the secondary circuit design.

Description shall be accompanied with the process flow diagrams and drawings (the SG installation in the compartments, the assembled SG, the main SG components, the SG EPS components).

1.9.3. Third circuit and steam turbine plant

Information on the third circuit systems, the steam turbine plant and the associated systems shall be provided. The information shall briefly reflect the configuration and boundaries of the steam turbine plant. At the same time brief information on interaction of the steam turbine plant and the RP both technologically through the parameters and via the CPS shall be presented.

Protection of the turbine set, the pipelines and the pressurized vessels against missiles that can cause any damage of the SS buildings or cable routes shall be demonstrated and substantiated.

Information on substantiation of strength, stability and operability of the steam turbine plant and the associated systems under any external natural and human-induced impacts shall be presented.

The process flow diagram and layout drawings (plans and sections) of the steam turbine plant shall be provided.

Design solutions for the third circuit and the steam turbine plant shall be analyzed.

1.9.4. Heat removal systems and ultimate heat sinks

Brief description of the systems shall be provided including:

- components for heat removal to the ultimate heat sink;

- service water supply sources;

- water circulation systems;

- the service water supply system.

The description shall contain the following: classification of systems, buildings, structures, the main thermal, hydraulic and structural characteristics of systems and equipment (supply and discharge channels, water intake structures, pump stations, cooling towers, makeup systems and sources for the circulating systems), the fundamental principles and criteria used for the design, operation modes, particularly in case of any operational occurrences, design basis accidents and external impacts.

Process flow diagrams shall be attached to the description.

1.9.5. Electrical systems

Brief description of the electrical systems shall be provided, particularly:

- configuration, purpose and classification of the systems;

- the grid connection diagram, number of the power supply lines, voltage in the lines;

- auxiliary power supply of the NPP power unit from external and internal sources;

- description of safety classes for the normal operation electrical systems and components;

- fire protection of the electrical installations;

- operation of the electrical systems in case of any operational occurrences, accidents, external natural and human0induced impacts.

The following principal schemes shall be attached:

the grid connection diagram for the NPP power unit;

the main electric circuit diagram;

the auxiliary power supply principal scheme;

the structural scheme of electrical protections;

the structural scheme of the CS;

the structural scheme of electrical systems.

1.9.6. Water chemistry regime

The concept of the WCR selection for the third circuit shall be stated.

It should be demonstrated that the WCR standards for the third circuit comply with the standards for steam turbine plants of the thermal power facilities.

1.9.7. Nuclear fuel storage and handling systems

The following data shall be provided:

- the list of all storage facilities for both fresh and spent nuclear fuel;

- characteristics of the fresh NF used at the NPP power unit as well as the fuel removed from the nuclear core with indication of the methods for the NF burn-up determination;

- the maximum design capacity (volume) of each NF storage facility and the number of places reserved for emergency core unloading and storage of rejected nuclear fuel (both fresh and spent);

- the method for the NF storage in FFSs and SFSs;

- the method for the NF delivery to the NPP and SNF removal from the NPP, information on the proposed transportation frequency and the TC types used;

- information on the in-plant transportation (means of transport and types of transport packages);

- information on the rejected NF handling (for both fresh and spent NF) beginning with the rejection method;

- the list of initiating events the set of NF (SNF) storage and handling systems is designed for and analysis of any emergency situations and design basis accidents.

1.9.8. Radioactive waste management

1.9.8.1. Liquid radioactive waste management

Brief characteristics of the RW handling systems, criteria and principles of their design shall be provided.

1.9.8.2. Solid radioactive waste management system

Brief characteristics of the system and the basic criteria and principles for its design shall be specified.

1.9.8.3. Gaseous radioactive waste management system

Brief characteristics of the system and the basic criteria and principles for its design shall be specified.

1.9.8.4. Gaseous radioactive waste collection and purification system

The active gas purification systems used to reduce releases of radioactive aerosols, various iodine forms (aerosol, steam and organic) and inert radioactive gases to the atmosphere and the NPP rooms shall be described. The decontamination factors shall be given for all systems separately.

1.9.9. Process control system

This subsection shall contain brief information on the NPP power unit CS, particularly on its structure, classification of the process control subsystems of the NPP unit, location of the CS room in the NPP unit building, control rooms of the NPP power unit, the warning and alarm annunciation systems for the power unit workers.

1.9.10. Safety systems

The list of protective, localizing, support and control safety systems and brief description of each system containing the following information shall be provided:

- purpose and configuration;

- compliance with the safety principles and criteria;

- criteria for performance of the prescribed functions by the system;

- brief description of the system: the flow diagram, arrangement, protection against internal and external impacts, monitoring and control;

- state of the system under normal operation conditions; integral testing and monitoring of the system;

- the system operation mode in case of any accidents.

1.9.11. General layout

The general layout shall be provided on the basis of topographic material with the scale of 1:2000 and the horizontal interval of 0.5 m, and in case of necessity - on topographic material with the scale of 1:500 and the horizontal interval of 0.25-0.5 m.

The list of the main buildings and structures of the power unit shall be provided together with the general layout.


- conditions determining the location of the main buildings and structures in the general layout (process interrelations, the natural terrain of the area, the prevailing wind direction, geological and hydrological conditions at the site, construction priority of the power units, etc.);

- orientation of the main NPP buildings;

- distances between the main buildings and structures and their justification;

- substantiation for location of the hydrotechnical structures, outdoor switchgears and auxiliary buildings and structures in the general layout;

- motor roads and railways, conditions for entrance into the main buildings and structures;

- terrain inclination at the site;

- grade elevations of the site;

- protection of the site against runoff;

- utility networks, transport, process and electrical connections between the buildings and structures, between the controlled and uncontrolled access areas.

1.9.12. Ventilation systems

Brief information on the main safety-related normal operation systems particularly the ones related to safety systems and radiation safety of the environment and the workers (if they are not described in Subsection 1.9.10) shall be provided.

The information shall include:

- purpose of each system;

- configuration of the system;

- design criteria;

- operation modes.

The schemes of the main ventilation systems classified as support ones (if they are not provided in Subsection 1.9.10) shall be attached to the description.

1.9.13. Radiation protection and radiological monitoring

General information on radiation protection for the main radiation sources specified in Section 10 and Subsection 11.4.3 shall be provided.

Criteria for selection of radiological monitoring hardware, development of the sampling point scheme and location of the equipment shall be specified. General description of the radiological monitoring hardware and the ARSMS shall be presented.

1.9.14. Physical protection system

Brief information on the PPS configuration and the requirements for the PPS shall be provided.

1.9.15. Fire safety arrangements

The brief list of the main fire safety arrangements for the rooms with typical fire load (cable rooms, rooms with the equipment of the primary and secondary circuit, rooms of the oil systems, control rooms, rooms of the instrumentation and control systems, rooms with computational equipment, etc.) shall be provided. The main fire safety arrangements (with regard to the construction, electrical and thermal-mechanical part) shall be stipulated for these rooms. Information on equipment of these rooms with automatic fire alarm and fire extinguishing units shall be presented.

The measures provided in the design to ensure evacuation of the workers in case of a fire and smoke protection of the buildings shall be described in brief.

The hydrogen safety system shall be described in brief.

Fire-fighting water supply of the production site, the main buildings and structures of the power unit, equipment of these buildings with the internal firewater pipeline shall be described; in this case fires determining design firewater flow rates shall be listed.

All water bodies and tanks that can be used for water intake to the mobile fire extinguishing equipment shall be listed.

Expected consequences of fires from the viewpoint of safety including expected consequences of fires in case of any building and structure destruction due to external impacts shall be specified.

The complete list of the fire detection and extinguishing systems as well as the emergency fire ventilation and smoke protection systems (system components) with indication of their safety class and seismic resistance in accordance with the federal rules and regulations in the area of atomic energy use shall be provided.

Fires occurring due to sodium coolant leakages from the primary and the secondary circuit shall be analyzed.

The design basis accident situation shall be defined for each room with the equipment of the primary and secondary circuit, and the maximum sodium spillage volume shall be specified. The potential fire parameters shall be calculated for the standard rooms: air temperature, temperature on the surface of walls, floors and ceilings, potential excessive pressure in the room (usage of the emergency fire ventilation shall be substantiated in case of necessity).

Sodium extinguishing methods shall be described for each room of the primary and secondary circuit, particularly:

- related to any risks for the operating power unit in case of a fire occurrence during construction of additional power units;

- the fire annunciation and communication system;

- arrangement of the fire-fighting service.

1.10. Brief description of the NPP power unit operation

- Preparation of the power unit for start-up

Brief information on the stages of preparation of the RP and the power unit systems for start-up shall be specified:

- state of individual elements and components of the RP and the power unit;

- filling of the primary and secondary circuit with sodium;

- the RCP start-up;

- leak-tightness and strength testing for the primary, secondary and third circuit;

- the ERCS testing;

- the SG EPS testing.

- The unit start-up from the cold state to full load

Brief information on the stages of the power unit start-up from the cold state to full load shall be provided:

- the method for the reactor nuclear core warm-up;

- monitoring of the nuclear core state;

- inspection of protections and interlocks;

- integral inspection of the CPS;

- the reactor power rising to the MCL;

- the RP power rising.

The following information shall be provided: parameters of the primary and secondary circuit coolant (P, t), pressure in the third circuit, the RP warm-up rates and completion conditions, the conditions for the reactor power rising to the MCL, the reactor power enabling connection of the turbine, the coolant parameters upon reaching of the rated RP power.

The start-up and loading schedule (changes of the basic parameters of the primary, secondary and third circuit, power rising from 0 to 100%) shall be provided.

- Power operation


- the range of power operation with due regard for accuracy of power maintenance by the control system;

- the operation mode.

- Adjustment of the unit power

Brief information on operation of the main controllers for the reactor and steam turbine plants shall be provided.

- Transient modes

The following information shall be presented for each design transient mode:

- scheduled shutdown of the primary and secondary circuit RCP (the regime development, the RP power reduction value depending on the number of disconnected RCPs, the procedure for the RCP and SG shutdown);

- connection of the previously idle loop (brief characteristics of the regime development, the RP power value prior to the RCP connection);

- scheduled shutdown of the TDFP (the regime development including: the initial state, preliminary power reduction, the RP power value depending on the number of operating TDFPs).

- The power unit shutdown to the hot state

Brief information on the regime development with description of the following cooldown stages shall be provided:

- "hot shutdown";

- sequence of operation for the primary, secondary and third circuit systems, cooldown rate;

- load reduction for the turbine generator, the RP power reduction, the basic controlled parameters;

- bringing of the reactor into the "hot state" with prior assurance of its sub-criticality;

- the method for cooldown and residual heat removal.

Information on the parameters of each cooldown stage for the primary, secondary and third circuit shall be provided.

The cooldown schedule shall be presented.

- Permissible maintenance works for the power unit in the "hot state"


- the coolant temperature and pressure with due regard for the brittle strength assurance conditions (in case these conditions are stipulated in the RP design);

- the possibility to eliminate any failure and to perform the RP maintenance in the course of "hot shutdown".

- The power unit cooldown to the cold state

Brief information on the regime development shall be provided including:

- definition of the "cold state" mode;

- the list of the main modes resulting in the necessity to bring the reactor into the cold state;

- sequence of operation for the primary, secondary and third circuit systems, cooldown rate;

- the method for cooldown and residual heat removal;

- the reactor sub-criticality and the methods to maintain it.

Information on the parameters of each cooldown stage for the primary, secondary and third circuit shall be provided.

The cooldown schedule shall be presented.

- Refueling

Brief information on the NF reloading regulations including operations for the SNF unloading from the reactor to the SFAD, rearrangement of NF in the nuclear core and loading of fresh NF as well as control of leak-tightness for the fuel elements shall be provided.

1.11. Environmental impact of the NPP power unit

Brief information on substantiation of the design solutions for chemical, radiation, thermal, electromagnetic and acoustic impact on the environment shall be provided.

The following shall be specified:

- Expropriation of certain territories, modification of the natural landscape, individual changes of the social and economic conditions in the NPP location area take place in the course of the NPP power unit construction. Environmental influence of the NPP shall be assessed differentially for each type of impact with due regard for the entire biosphere diversity, i.e. the influence of each type of impact on the eco-system, biota, flora, fauna and the humans.

- Environmental impact of the NPP power unit shall be assessed with due regard for its actual state, the ecological situation in the NPP location area, the existing sanitary and hygienic, biological, anthropogenic and human-induced characteristics of the biosphere pollution.

- Results of the integral environmental impact assessment for the NPP power unit shall be provided.

1.12. Comparison of the NPP power unit design with similar design of national and foreign NPPs

The selected counterparts of the NPP power unit design shall be defined.

A power unit with the same RP type where the same or similar principle of safety assurance, control and protection is implemented may be considered as the counterpart of the NPP power unit.

In the absence of any suitable counterpart the NPP power unit may be compared with NPP power units having the same reactor type similar in the rated capacity.

The presented NPP power unit design should be compared with the counterpart mainly with regard to safety systems. It is recommended to provide the layout drawings of the presented design and the counterpart design on the scale of 1:500 and the principal schemes of the counterpart for comparison.

1.13. The NPP power unit construction schedule, contracting parties and contractors

The network schedule of the power unit construction, the names of all participants (the design developers and the construction contractors) as well as the information on the operating organization, contracting parties and contractors engaged in the construction shall be provided.

1.14. Fundamental provisions for the NPP power unit operation arrangement

1.14.1. Commissioning of the NPP power unit

Brief information on the commissioning work program including testing of the structures, systems and components in the course of the NPP power unit commissioning shall be provided.

The main stages of the commissioning works shall be listed with description of their plans enabling to assess the possibility for successful performance of the commissioning works and success criteria for completion of all items mentioned in the plan. The objective to be achieved in the course of inspections and tests shall be specified for each stage.

The procedure for issuance and storage of the reporting documentation with indication of conditions for access thereof shall be described.

1.14.2. Management of the NPP power unit operation

Information on preparation and arrangement of the power unit operation shall be provided including brief description of the operating organization structure with the focus on the responsibilities of individual persons and departments for the power plant operation. Description of the operating organization shall include the basic issues on training of the NPP personnel with the required qualification (availability of training centers, training programs, timely training, the procedure for competence assessment and admittance to unsupervised works).

Efficiency of the actions for maintenance and monitoring of the power plant operational state shall be demonstrated. In particular, it should be demonstrated how the inspection and testing results are taken into account in the NPP safety assessment programs, how the operation experience is considered in development of the maintenance schedule; the procedure for preparation and submittal of the regular information on the current safety level shall be also described.

1.14.3. Safe operation limits and conditions

General information on the safe operation limits and conditions shall be provided.

The safe operation limits and conditions established in the design for the power unit are given in Section 16.

1.14.4. Decommissioning of the NPP power unit

The main provision of the NPP power unit decommissioning concept shall be stated.

Detailed information on the NPP power unit decommissioning is presented in Section 18.

1.15. Quality assurance

Brief information on the activities of the NPP power unit construction participants confirming the ability of these organizations to ensure quality of the performed works and rendered services affecting the NPP safety shall be provided.

Description of the quality assurance system demonstrating interaction between the operating organization, the NPP design developers and any organizations providing services in the area of atomic energy use shall be presented.


- responsibility of each organization for the NPP power unit construction quality assurance;

- availability of independent quality assurance control in the operating organization for all works, products or services affecting the NPP safety;

- progress of the quality system development, implementation and functioning in the operating organization and other nuclear facilities;

- progress of the quality assurance program development and implementation in the operating organization and other nuclear facilities as of the NPP SAR submittal moment.

1.16. General assessment of the design

It shall be stated that the NPP power unit is designed (under design), constructed and will be operated in accordance with the requirements of federal laws and regulatory documents in the area of safety.

2. REQUIREMENTS FOR THE SECTION "CHARACTERISTICS OF THE NPP LOCATION AREA AND SITE"

Characteristics of the NPP location area and site shall be presented, and acceptability of the siting conditions for the NPP shall be assessed.

Information on the environmental conditions at the NPP site as well as on any conditions related to human activities, the existing and expected population distribution and land use shall be provided.

Completeness and adequacy of the surveys and studies aimed to obtain credible characteristics of the NPP location area and site shall be substantiated.

The following parameters and characteristics shall be provided:

- external natural and human-induced impacts on the NPP power unit;

- environmental impact of the NPP.

Impacts shall be taken into account in the subsection "Design basis" for the buildings, structures and process systems with regard to emergency planning including planning of the public evacuation in case of emergency.

The effective regulatory documents shall be followed to determine the parameters and characteristics of external impacts on the NPP power unit and the environmental impact of the NPP power unit.

The scope of information on the selected and approved site shall comply with the requirements of this RD.

2.1. Description of the NPP location area

The following territory coverage radii shall be adopted taking the main building (the RB) as the center:

- region - at least 300 km;

- locality - at least 30 km;

- site - at least 3 km.

2.1.1. Geographic location

The NPP location shall be recorded with indication of the latitude, longitude and altitude in the unified coordinate and elevation system.


- administrative location of the site;

- distance to the administrative center and its name;

- distance to the nearest administrative boundaries;

- distance to the national boundaries and names of the nearest states;

- position of the site in relation to natural and artificial benchmarks (populated localities, rivers, seas, airports, railway stations, sea and river ports, etc.);

- industrial and military facilities (if any) with due regard for their potential expansion;

- distance to any recreational areas, national parks, restricted areas, etc.;

- transport facilities and utility systems.

The list of industrial, military, transport facilities, process pipelines and any other elements that can affect the NPP power unit safety or be affected by the power unit operation shall be presented.

2.1.2. Topographic conditions

The list of materials with the results of engineering and geodetic survey and analysis of these results shall be provided.

The following shall be specified in the description of the area and site terrain:

- maximum and minimum absolute elevations;

- surface inclination and its direction;

- peculiar terrain details;

- boggy areas;

- forests, fields and other agricultural lands.

Topographic and geodetic materials shall be presented in the uniform coordinate and elevation system.

The following materials shall be provided for the territory within the radius of at least 30 km from the main building:

- the topographic map on the scale of 1:25 000;

- the topographic and bathymetric plan of the shelf zone and the map on the scale of 1:10,000 with the bottom horizontal interval of 5-2.5 m;

- materials on observations over present-day crustal motion;

- the topographic map (plan) of the aite on the scale of 1:5000;

- the topographic and bathymetric plans (maps) of the shelf zone at the site on the scale of 1:5000.

The topographic materials shall be obtained not later than five years prior to their presentation.

2.1.3. Demographics

Demographic data shall be based on the results of the latest census survey with due regard for the expected migration and changes of the demographic situation within the NPP area for the entire operation period.


- the population density within the radius of 30 km from the NPP prior to commencement of the construction, during the construction period and for the entire NPP operation period;

- distance to the cities with the population size exceeding 100 000 people for the zone with the radius of 100 km from the NPP;

- distribution of the population by sectors (circles) around the NPP confined with the radius of 10 km and divided into 8 rhumbs;

- information on any expected changes of the population density within the entire NPP operation period including seasonal and daily density changes.

2.2. Hydrometeorological conditions

2.2.1. Climate


- wind direction and speed, the wind rose;

- average and extreme values of air saturation with water vapors (absolute and relative humidity);

- average and extreme quantity of precipitation (rain, snow), duration of precipitation, precipitation distribution according to intensity and monthly diagrams of winds bringing the precipitation;

- average and maximum repeatability and duration of fogs, smogs, thunderstorms, blizzards, hail, glaze, dust and sand storms;

- average and extreme air temperatures;

- average and extreme temperature of soil on the surface and at standard depths;

- average and extreme atmospheric pressure;

- pollution, dust content and corrosive activity of the atmosphere;

- chemical composition and description of the surface water ability to disperse, dilute or concentrate wastes;

- probability assessment for hazardous hydrological and meteorological phenomena (tornado, cyclone, snow avalanche, tsunami);

- aerological conditions (repeatability of no-wind conditions and wind directions; average wind speeds in 16 rhumbs at the height of 100 and 200 m; average vertical temperature gradient in the layers of 0-300, 0-600 and 0-900 m; repeatability and average thickness and intensity of ground inversions; repeatability and average thickness and intensity of raised inversions in the layer of 0-2 km; atmospheric stability; atmospheric dispersion of impurities).

2.2.2. Meteorological and hydrological conditions at the site

Analysis results for the meteorological and hydrological conditions including the list of hydro-meteorological processes and phenomena detected within the NPP location area and conclusion on presence or absence of any processes and phenomena at the NPP power unit site shall be provided.

Information on each process and phenomenon shall be presented separately. Their intensity and frequency shall be substantiated with the results of special-purpose observations, calculations and analysis of statistical data.

2.2.3. Basic materials

The list of materials used to determine characteristics and parameters of hydro-meteorological impacts shall be provided, particularly:

- historical data from chronicles, archives and newspapers as well as photographs;

- witness reports;

- hydro-meteorological observation data for the NPP power unit site in accordance with the standard methodologies;

- sets of annual parameter values and information on any significant maximums on the long-term scale (up to 50 years);

- design probability and parameters of impacts.

2.2.4. Calculation methods and results

Methods and results of the calculation of parameters and characteristics of the hydro-meteorological processes and phenomena listed below shall be described.

2.2.4.1. Wind

- Design wind speeds in vertical direction, wind repeatability intervals and blast coefficients.

- Description of the methods for conversion of wind speed into effective pressure on the structure surfaces facing the wind.

- Results of the wind load calculations, mode factors for the structures, wind pressure distribution along the structure height.

2.2.4.2. Tornado

- Input data for calculation of loads caused by a tornado:

forward speed;

tangential speed;

pressure differential;

characteristics of building or structure fragments and other missiles caused by a tornado.

- Coefficients of pressure pattern and distribution on flat surfaces and round structures (like the containment).

- Combination of loads under the most unfavorable impact of tornado on the structure.

2.2.4.3. Extreme snowfalls and snowpacks

- Maximum height of snow cover on a horizontal surface.

- Snow load distribution patterns.

-Factors of snow blanket mass conversion into snow load on the surface.

2.2.4.4. Glaze frost

- Calculation of the rated linear ice load on the elements of circular section.

- Calculation of the rated surface ise load on any other elements.

2.2.4.5. Air temperature

- Calculation of change in average temperature and temperature differential over time in warm and cold seasons.

- Calculation of average daily external air temperatures in warm and cold seasons.

- Temperature increment calculation.

- Calculation of the initial temperature in warm and cold seasons.

2.2.4.6. Snow avalanche

- Calculation of static and dynamic pressure of sliding snow on the snow-retaining structures.

- Calculation of the avalanche impact strength per 1m2 of a fixed stiff obstacle surface located perpendicular to the avalanche movement direction.

- Calculation of the avalanche load on the damping obstacle during avalanche flow around it.

- Calculation of pressure in case of a diagonal avalanche impact.

- Calculation of loads on the facility roof.

- Calculation of avalanche pressure on the concave surface.

- Calculation of overpressure at the ASW front.

2.2.4.7. Flood

- Absolute elevation of the site flooding level.

- Water flow velocity.

2.2.4.8. Negative and positive setup, storm waves in the coastal area

- Absolute elevation of the territory flooding level.

- Flooding area.

- Dynamic impact of flooding caused by a storm.

2.2.4.9. Tsunami

- Wave height.

- Water setup and recession height.

- Dynamic impact of the tsunami wave.

2.2.4.10. Seiches


2.2.4.11. Extreme precipitation

- Height of the precipitation layer.

2.2.4.12. Tides


- Absolute elevation of the coastal area dereliction level.

2.2.4.13. Ice blocks and jams in the water courses


- Dynamic impact of flooding caused by ice blocks and jams.

2.2.4.14. Extremely low flow, abnormal water level decrease

The impact of the water level increase or decrease at the site shall be assessed for the phenomena specified in par. 2.2.4.7 - 2.2.4.13. In this case:

- the possibility of flooding shall be substantiated based on the calculation of the water level during flood and (or) rise of the groundwater level;

- calculations of the high water level, the peak water flow rate due to precipitation, floods, seiches, tsunami, waves, ice jams, tides, water reservoir breakdowns shall be presented;

- calculations of the water level decrease due to heavy drought, seiches, tsunamis, waves, ice jams, negative setup, ebbs and other phenomena shall be presented;

- phenomena considered in the NPP power unit design shall be specified, and characteristics of their impact on the structures and systems of the NPP power unit shall be described.

2.3. Geological, tectonic, hydrogeological, seismic and geotechnical conditions

The materials of engineering and geological survey and investigation of the geological and tectonic structure, the recent tectonics, seismotectonics and seismicity in the NPP location area shall be provided. The list of hazardous geological processes and phenomena specified in the RD shall be presented in accordance with the nomenclature. The methods for calculation of the main parameters for the geological and seismic processes and phenomena shall be specified. Predictions of any unfavorable changes in geological, hydrogeological and seismic conditions that can trigger any geological hazard during the construction, operation and decommissioning of the NPP power unit shall be provided. Information on the soil stability and properties shall be specified.


The list of materials containing the consolidated (summary) results of surveys and investigations of the geological and tectonic structure, the recent tectonics, seismotectonics and seismicity in the NPP location area as well as hydrogeological, geotechnical and seismic conditions at the NPP power unit site shall be presented.

2.3.2. Results of analysis

Analysis results for the basic materials with conclusions on presence or absence of any geological hazard at the NPP site and the list of detected geological hazards shall be provided; the characteristics and parameters of the processes subject to consideration in the NPP design and operation shall be described.

Information on each type of processes and phenomena shall be presented in the following order:

- Fissure displacements, seismic dislocations, seismic and tectonic upheavals and settling of crustal blocks.

- Modern differential crust movements including tectonic creep.

- Residual seismic deformations of the crust.

- Earthquakes of any genesis.

- Volcanic eruption.

- Mud volcanism.

- Landslides of any genesis.

- Rockfalls and earth slip-falls.

- Mudflows.

- Snow and stone and crushed gravel-block avalanches.

- Scouring of banks, slopes, stream beds.

- Sinks and subsidences

- Underground erosion including karst formations.

- Freeze-thaw (cryogenic) geological processes.

- Deformations of specific soils.

- Micro-deformations of soils at the foundations of essential structures of the NPP power unit.

Any possible combinations of interrelated processes and phenomena shall be considered.

The hazard degree of the processes and phenomena, their intensity and frequency of occurrence shall be assessed. Estimates and predictions shall be substantiated with descriptions, graphic and numerical materials (profiles, plans, sections, core samples, maps, photographs), results of their analysis, as well as special-purpose field and laboratory studies.

2.3.2.1. NPP location area


- Analysis results for the archive and library materials on the main directions of engineering surveys and investigations.

- Cartographic schemes and profiles (on the scale of 1:100 000 - 1:500 000) on geology, tectonics, the recent and modern motions, particularly the seismotectonic map or the geological seismicity criteria map, the detailed seismic zoning map, the schematic map of potential earthquake source zones with indication of the expected maximum magnitude, its repeatability and the effective focal depth in each zone; historical data on earthquakes and any other geological and geotechnical processes.

- Description of lithology and stratigraphy in the region, composition and thickness of quaternary deposits, structure and occurrence depth of the basement rock.

- Schematic maps of zoning in accordance with the hazard of exogenous geological processes.

- Soil freezing depth, thickness of the active layer.

- Information on landslides, rock falls, subsidences and depressions, karst and creek formation, scouring of banks.

- Predictions of any potential soil deformations due to extraction of gas, liquid and solid mineral resources and as a result of any human-induced loads on the ground surface (water reservoirs, dense multi-storeyed development, seismic effects of quarry explosions, etc.).

- Observed settlement and tilt of the foundations for buildings and structures.

- Geodetic observation data for the present-day crustal motion.

- Information on the hydrogeological conditions:

groundwater depth and level variations;

interconnections of the aquifers and their connections to the surface water;

aquifer makeup and discharge areas;

assessment of hydrogeological dispersion in groundwater;

groundwater level depth with the probability of 10% and seasonal level variations, flow directions and velocities, soil filtration coefficients shall be indicated on the hydrogeological maps.

- Results of macro-seismic and instrumental seismological studies in the region.

- Description of soil categories in accordance with seismic properties and their location at the NPP site.

- Geological and geophysical profiles and block schemes of the main key strata up to the depth of 100-300 m on the scale of: horizontal 1:100 000 - 1:500 000, vertical 1:5000 - 1:20 000 (the scales within the region: horizontal 1:20 000 - 1:50 000, vertical 1:1000 - 1:5000).

- Interpretation results for aerial and satellite photographs.

2.3.2.2. NPP site


- Geotechnical zoning and seismic micro-zoning maps for the site with indication of the geological profiles, reference wells and the main structures (the scale: horizontal 1:2000 - 1:10000, vertical - 1:200 - 1:1000).

- Geotechnical cross-sections, cores of exploratory holes drilled in the areas for location of essential structures and additional profiles along the axes of the essential structures (the scale: horizontal 1:500 - 1:2000, vertical 1:50 - 1:200). All layers (engineering-geological elements) shall be distinguished and described on the profiles, the rated physical, mechanical and dynamic properties of soils in their natural and water-saturated state (and in natural and unfrozen state for permafrost soils) under dynamic impacts and static impact from the weight of structures shall be specified. Presence of any lenses and interlayers of detrimental soils with unstable properties in the profile shall be specifically noted. Recommendations for improvement of soil properties shall be provided.

- Seismic characteristics of the site:

intensity for the medium grade of soils according to the MSK-64 scale;

SSE and DBE at the site with due regard for human-induced changes (leveling, dewatering, flooding of the territory, etc.);

design accelerograms and consolidated soil response spectra in graphical and digital format.

- Geodynamic characteristics of the site.

2.3.3. Survey and investigation methods

Ii is necessary to provide information on the methods, methodology, hardware and equipment used:

- for seismic survey and electrical exploration as well as other geological and geophysical investigations;

- for determination of physical and mechanical soil properties, specific properties of subsidental, swelling, fluid and fluid-plastic, weak and permafrost soils in each layer of the compressed strata, chemical composition of groundwater.

Accuracy characteristics for the equipment, units and methods shall be specified.

2.3.4. Prediction methods

Methods for prediction of characteristics and parameters of processes and phenomena shall be described.

2.4. NPP siting conditions related to human activities


Text, tables, maps, schemes related to external human-induced events shall be sufficient to assess probability of these events and to predict the parameters and characteristics of the associated impacts. The following information shall be provided for the events listed below.

2.4.1.1. Crash of aircraft or other missiles

- Location of airports and position of air corridors and intersections of air routes shall be indicated on the general map of the region.

- Kinds of air traffic, types of aircraft and their characteristics, frequency of flights.

- Presence of any military facilities particularly a bombing ground at the distance of up to 30 km from the NPP.

- Types of potential missiles, their characteristics, frequency of occurrence related to human activities.

- Archive data on airplane crashes.

2.4.1.2. Fire due to external reasons

1. Potential fire sources shall be indicated on the general map of the region:

- forestation;

- warehouses for solid, liquid and gaseous explosive substances;

- main oil, gas and product pipelines;

- motorways and railways, navigable rivers and sea routes;

- air fields, air route corridors;

- residential areas;

- industrial facilities;

- coal and peat extraction facilities;

- areas with peat deposits.

2. Archive data on fires in the region

3. Information on the stocks of flammable materials in the fire hazard sources.

4. Wind rose

2.4.1.3. Explosions at facilities

- Distance from the NPP to any stationary and mobile potential explosion sources including:

warehouses, storage facilities, transportation vehicles with explosive substances;

pressurized vessels and units with gases or hot liquids;

buildings, structures and enterprises where hazardous technologies are used and internal explosions are possible;

motorways and railways, water transport with indication of the data on transported explosive substances;

main oil, gas and product pipelines;

military facilities.

- Information on the stocks of explosive substances.

- Archive and statistic data on any explosions in the region.

2.4.1.4. Water reservoir breakdown

- Plan of the water reservoir location in relation to the NPP.

- Resistance of the hydrotechnical structures to external impacts of natural and human-induced origin.

- Hydro-meteorological data on the long-term scale (at least 50 years) including sets of annual values of the parameters as well as information on any extreme events.

- Data on the annual water level measurements in the upstream reach.

- Estimates of the maximum water storage in the upstream reach.

2.4.1.5. Corrosive impact

- Results of chemical analysis for water and soil samples in the region.

- Brief description of the hydrogeological conditions at the site: characteristics of the aquifers, chemical composition of groundwater, groundwater level variations, any potential flooding of the underground NPP structures, conditions for the perched layer formation; aggressiveness of soils below the groundwater level.

- Probability assessment for any releases of corrosive substances stored, produced and transported within the region.

2.4.1.6. Releases of explosive, flammable and toxic vapors, gases and aerosols into the atmosphere

- Distance from the NPP to any industrial facilities using chlorine, hydrogen sulfide, ammonia, sulfur dioxide and any other active substances, places of chemical releases.

- Schemes of movement of mobile toxic hazard sources

2.4.1.7. The list of external impacts of human-induced origin considered in the design

The list of external impacts of human-induced origin considered in the design

Impacts with very low probability or insignificant intensity and (or) impacts from remote sources may be neglected. Safe distances and intensity values for individual types of impacts shall be determined in accordance with special standards.

2.4.2. Calculation methods

Methods and methodologies for calculation of the basic parameters and characteristics of human-induced external impacts shall be described.

2.4.3. Prediction results

The following parameters and characteristics of external impacts shall be assessed for the events listed below.

2.4.3.1. Crash of aircraft or other missiles

- Probability of the event.

- Stiffness properties and weight of colliding bodies.

- Impact direction and velocity.

- Collision angle and area.

2.4.3.2. Fire due to external reasons

- Probable area of the territory affected by fire.

- Heat flux in the source of fire and its changes towards the NPP.

- Distance to the NPP.

- Wind speed and direction assumed for the calculation.

2.4.3.3. External explosion

- TNT equivalent of the explosion and its distance to the NPP power unit.

- Overpressure at the ASW front.

2.4.3.4. Releases of explosive, flammable and toxic vapors, gases and aerosols into the atmosphere.

- Quantity of the substance that can be involved in the event.

- Initial concentration at the release point; dispersion of the releases in the atmosphere; concentration due to the primary sources and secondary effects; duration of exposure.

- Wind speed and direction assumed for the calculation.

- Presence and capacity of the ignition source (vapors, gases and aerosols).

- Concentration of the released cloud upon approach to the NPP.

2.4.3.5. Water reservoir breakdown

- Wave height and speed.

- Absolute elevation and duration of the NPP site flooding.

2.4.3.6. Electromagnetic pulses and emission

- Distance from the emission source to the NPP power unit.

- Intensity of electrical and magnetic fields.

2.5. Impact of the NPP power unit on the environment and the public

Information on the region required to assess the environmental impact of the NPP as well as the data on radioactive, chemical and thermal pollution of the environment shall be provided. Data on the concentration of radioactive products that can get into the human organism shall be specified.

Consequences of radionuclide releases and discharges into the environment in the course of normal operation shall be assessed.

Methods for monitoring of the environment and methods for determination of the "zero radiation background" shall be described.

2.6. Programs for observation over processes and phenomena within the period of the NPP design, construction and operation

2.6.1. List of programs

The list of programs for observation over processes and phenomena within the period of the NPP power unit design, construction and operation shall be specified to provide the following:

- studies and investigations of the present-day motions of the near-surface crustal layers at the depth of the foundation beds for the NPP structures;

- observations over micro-deformations on unstable slopes and at the bases of the NPP structures;

- definition of the characteristics and geometric parameters of any tectonic abnormalities detected within the region, locality and the NPP site subsequent to analysis of the library and archive materials;

- seismometric measurements of earthquakes and vibrations caused by explosions;

- observations over the groundwater regime;

- observations over the surface water regime;

- meteorological observations;

- measurement of the soil humidity, density and load-bearing capacity (geotechnical control);

- observations over earth slips, development of karst funnels and other processes and phenomena in the NPP location area.

The programs shall be provided for each type of observations.

2.6.2. Description of programs

The following shall be specified and described in each observation program:

- the list of observed conditions, processes and phenomena;

- the list of observation types;

- location of the measurement points;

- performance of measurements;

- measurement methods, characteristics of the equipment and test facilities (references to par. 2.3.3 may be given);

- contents of the monitoring report.

2.7. Support of the workers and the public and their evacuation in case of any emergency

Analysis results for emergency situations at the NPP power unit and in the NPP location area caused by an intensive earthquake, any other extreme external impacts and combination thereof as well as action plans for emergency situations shall be provided. Administrative and technical measures for evacuation, particularly in case of destruction of any transportation lines, air fields, bridges, tunnels due to depression, overthrust fault, ground surface fracture, landslides, rock falls and rock slides shall be described.

Recommendations for the use of the existing access routes in case of any emergencies, relocation and refurbishment of roads, bridges, ports, etc., construction of new transport routes for access to the NPP in three-four directions shall be presented.

2.8. Summary table of external impacts on the NPP

The following shall be specified in the summary table of external impacts on the NPP:

- characteristics and parameters of hydro-meteorological phenomena;

- characteristics and parameters of geodynamic, seismotectonic, geological, hydrogeological, seismic and geotechnical parameters, processes, phenomena and events;

- characteristics and parameters of human-induced impacts.

The approximate format of the table is given below.

No. Process, phenomenon, event

Source of the process, phenomenon or event

Degree of hazard

Frequency of occurrence

Impact parameters

Additional information

Natural processes and phenomena and external human-induced events considered in the NPP power unit design shall be presented in the table.

The list of initiating events considered in the action plans for protection of the workers in case of any emergencies shall be provided.

2.9. Documentation of the data on the NPP power unit siting conditions

The subsection shall be issued as the appendix to Section 2 and shall include information characterizing the NPP siting with regard to environmental conditions, processes, phenomena and external human-induced events affecting the NPP.

The subsection shall be compiled in such a way so that to enable registration of any changes in the siting conditions at all stages of the NPP lifecycle.

It is recommended to document the information on the NPP power unit siting conditions in accordance with the format presented in Appendix 2, beginning from the PSAR and FSAR development stage, and adjust this information in the course of the power unit operation.

3. REQUIREMENTS FOR THE SECTION "GENERAL PROVISIONS FOR DESIGN OF BUILDINGS, STRUCTURES, SYSTEMS AND COMPONENTS"

3.1. Basic principles and criteria for design of buildings, structures, systems and components

3.1.1. List of applied rules and standards

The list of effective federal rules and regulations in the area of atomic energy use applied by the Applicant in the course of design shall be provided.

3.1.2. Assessment of compliance with the requirements

Information on compliance with the basic principles of safety assurance for the NPP power unit shall be presented, particularly:

- compliance with the defense-in-depth principle, application of the barrier system in the way of ionizing radiation and RSb propagation into the environment, implementation of the system of administrative and technical measures including the accident management measures;

- approbation of safety-related design solutions by experiments and studies;

- quality assurance measures at all stages of the NPP lifecycle;

- approach to the human factor consideration aimed to eliminate errors or to mitigate consequences related to actions of the NPP workers, particularly in the course of maintenance;

- measures aimed to ensure non-exceedance of the established standards for release and discharge of radioactive substances into the environment;

- fire protection measures;

- organizational solutions for physical security assurance;

- measures for qualification and psychological preparation of the operating organization

employees in order to ensure compliance with the safety culture principle in the course of design.

3.1.3. Deviations, their substantiation and implemented compensatory measures

The list of deviations from the requirements of federal regulations and rules in the area of atomic energy use, substantiation of these deviations and the implemented compensatory measures shall be provided, and the reference to the document section (the list of deviations) where these deviations are justified in detail shall be given.

3.2. Applied classifications of buildings, structures, systems and components

3.2.1. Classification of the buildings, structures, systems and components in accordance with their safety importance

Information on classification of safety-related buildings, structures, systems and components into safety classes in accordance with the GSP shall be provided.

3.2.2. Classification of equipment and pipelines into quality groups

Information on classification of safety-related components into quality groups in accordance with the NPU Rules shall be provided.

3.2.3. Seismic classification

Information on classification of the buildings, structures, systems and components according to their seismic resistance in compliance with the Seismic Design Rules for Nuclear Power Plants shall be provided. The results shall be presented in the tabular format (Table 3.1).

Data specified in Column 7 shall be obtained from the analysis performed in Subsection 3.4.

3.2.4. The list of buildings, structures, systems and components subject to analysis of resistance to natural and human-induced impacts

Necessity of the analysis for resistance of the NPP buildings, structures, systems and components to natural and human-induced impacts in accordance with the requirements of the RD "Consideration of External Impacts on Nuclear and Radiation Hazardous Facilities" shall be indicated in Table 3.1 (Column 7).

Table 3.1

List of the NPP buildings, structures, systems and components and their classification

Designation of the building, structure, system and component

Name of the building, structure, system and component

Purpose (classification according to the purpose)

Safety class

Quality group

Seismic category (sub-category)

Consideration ofhuman-induced and naturalimpacts (results of probabilisti

c scenario analysis)

1 2 3 4 5 6 7

3.3. Description and substantiation of the layout of buildings and structures

The general layout of the NPP, its description and substantiation of the territorial location of buildings and structures in order to ensure the NPP safety under all natural and human-induced impacts considered in the design shall be provided.

Location of the water supply lines, communication lines, access ways, water intake units, switchgears, surface and underground storage facilities for diesel fuel and oil, the transformer yard, storage facilities for fire- and explosion-hazardous substances shall be indicated on the general layout of the NPP.

Brief description and substantiation of the location, dimensions and basic engineering and technical solutions shall be provided for the following buildings and structures.

1. Main building:

- the RB including the rooms of the emergency reactor cooldown system and other SSs and CSs.

- Steam generator compartment.

- Turbine generator compartment with the deaerator room and electric equipment racks.

- SNF compartment.

- Plenum ventilation center.

- Electric equipment compartment.

2. Special building.

3. SDGS.

4. Service water pump station.

5. MCR room.

6. Supply and discharge channels (service water, cables and other SRS networks).

7. Spray cooling pond or cooling tower (if any).

8. RW repository or warehouse.

9. Demineralized water storage tank.

10. Foundation bed of the main building.

11. Fire extinguishing pump station.

12. Accident management center (if any).

13. NPP buildings, structures and fencing related to physical security of the NPP.

Safety-related systems located in these buildings and structures shall be listed.

Fire protection measures (location of the buildings and structures on the general layout of the NPP) shall be described.

3.4. Probable development scenarios for initiating events of natural or human-induced origin at the NPP site

Results of review and qualitative analysis of the probable development scenarios for initiating events at the NPP due to the causes listed below shall be provided in accordance with the requirements of the RD "Consideration of External Impacts on Nuclear and Radiation Hazardous Facilities" :

- natural and human-induced external impacts;

- impacts caused by accidents at the NPP site.

For convenience of the analysis it is recommended to present the scenario review results in the table according to the approximate format given in Appendix 3.

3.5. Parameters of the impacts caused by emergency situations occurring at the NPP site

3.5.1. Impacts caused by emergency situations at the NPP site outside the main building

3.5.1.1. Mechanical impacts:

- Air shock waves

Potential sources and causes of explosions resulting from breakage of pressurized vessels, tanks with liquefied or compressed gas, fires and explosions in the storage facilities for petrol, oil and lubricants, etc. shall be described and analyzed. Parameters used as the input data for calculation of the ASW impact shall be provided.

Methods used to calculate the ASW parameters, to convert the shock wave parameters into effective loads on the buildings and structures (references to Section 2 may be given) and to calculate dynamic loads from missiles caused by the ASW shall be described.

- Missiles

The possibility for formation of missiles due to development of any accidents including the ones caused by destruction of pressurized equipment with rotating parts because of the rotation speed exceedance or any accident at the units of high-pressure systems shall be analyzed.

Selection of particular missiles shall be substantiated. Missiles that can be formed in case of destruction of buildings, structures, warehouses with materials, storage facilities for liquefied or compressed gas, pipelines and any other equipment located at the NPP site shall be taken into account. The size, weight, energy, velocity and other parameters required to determine their penetration capacity shall be defined for the selected missiles. The areas of potential hitting with missiles (target areas) shall be indicated on the plans and profiles of the buildings and structures.

Mathematical models used to analyze formation of missiles and to determine their characteristics and flight paths shall be described.

3.5.1.2. Chemical and corrosive impact

Chemical composition and pH value of the media flowing and contained in the equipment and pipelines subjected to potential breakage shall be specified.

3.5.1.3. Impact of toxic gases and aerosols

Probability of any toxic gas and aerosol releases to the atmosphere due to any emergency situations shall be analyzed. Assessment methods and values of the toxicity parameters for these emergency situations shall be described.

Probability of gas and aerosol ingress to the rooms shall be analyzed, and safety of the workers shall be assessed.

3.5.1.4. Radiation impacts

In case any damages of buildings and (or) structures containing radioactive substances are possible due to emergency situations at the NPP site radiation intensity as well as parameters of radionuclide dispersal processes in the atmosphere, surface and ground water shall be defined.

3.5.1.5. Fire load

It should be explained in brief how the fire load is formed in case of any fires in fire-hazardous rooms with sodium, oils, in cable and other rooms, and what load combinations it can contribute to. It should be demonstrated for what structures safety factors shall be substantiated in consideration of the fire loads. Results of review and analysis shall be presented in the relevant sections of the NPP SAR.

3.5.2. Impacts caused by emergency situations in the main building

3.5.2.1. Mechanical and thermodynamic impacts:

- Air shock waves

Information shall be presented within the scope at least equal to the scope specified in par. 3.5.1.1.

- Missiles

Information shall be presented within the scope at least equal to the scope specified in par. 3.5.1.1. Analysis results for impacts of missiles formed due to destruction of the equipment installed inside the containment system (systems) on the integrity of the lining and other containment structures shall be presented.

- Missiles formed due to destruction of turbines

References to the information specified in Section 6 may be given.

а. Turbine location and orientation

The turbine location and orientation shall be indicated on the TP layout drawing (scheme).

The missile ejection areas with the size of +/-25 degrees in relation to the disks of low-pressure cylinders shall be specified on the turbine hall plan and profile.

The places of potential hitting with missiles (target areas) shall be indicated on the plan and profiles in relation to the safety-related normal operation systems.

b. Determination of the missile characteristics

Such characteristics as weight, shape, cross-section areas, turbine destruction rate as well as critical ejection angles of the missiles shall be included into the description of potential missiles formed

by the turbine destruction.

Mathematical models used to analyze formation of missiles, the turbine shell breakage and the flight path of missiles shall be described.

c. Probabilistic analysis

Analysis results with regard to probability of any missiles hitting the unit systems shall be presented, and brief description of the calculation methods shall be provided.

All assumptions used in the analysis shall be specified and input data for these assumptions shall be substantiated.

- Dynamic impacts caused by pipeline breakages

Description and classification of all possible impacts on the NPP structures, systems and equipment caused by any pipeline breakages shall be provided.

Schemes of the high- and medium-pressure pipeline routes with indication of the safety-related systems, equipment and structures located in the close vicinity of the pipelines shall be presented.

In case any accident in high- or medium-pressure pipelines results in steam exposure on the nearest safety-related structures or steam ingress into any other rooms and compartments of the building the impact of steam on the operation of the affected equipment, structure or system shall be analyzed, and the permissible limit conditions for their further operation shall be determined.

The high- and medium-pressure pipeline breakage points where fencing or safe location cannot be used shall be indicated, and places of the occurring load application to the equipment, structures and other systems and components shall be determined. Criteria for definition of the breakage and leakage points in the pipelines shall be provided.

The possibility for formation and impact of secondary missiles in these systems shall be analyzed.

Methods used to determine the functions required for dynamic analysis of the pipeline whip caused by their partial or complete breakage shall be described.

Description shall include direction, thrust coefficients, acceleration time, magnitude, duration and initial conditions sufficiently characterizing the jet stream dynamics and pressure drop in the system.

Mathematical modesl sued for dynamic analysis of response shall be provided. All dynamic factors used in the calculations shall be specified and substantiated.

Methods used to assess the jet impact and the load on the systems and equipment caused by the pipeline breakage or occurrence of a hole shall be presented. Analytical methods for strength checks of the equipment subjected to any load occurring in case of pipeline breakages shall be additionally provided.

Protective devices on the pipeline penetrations through the civil structures shall be described in brief.

3.5.2.2. Chemical and corrosive impact

Reactions of sodium interaction with the equipment materials, concrete and insulating coatings and paints shall be considered; toxicity, flammability, explosion hazard, chemical and corrosive activity of the products of these reactions shall be assessed. Corrosive damage levels for the materials of safety-related equipment, structures and structural assemblies shall be determined on the basis of this assessment; it should be demonstrated that these levels do not exceed the permissible limits.

3.5.2.3. Impact of toxic gases and aerosols

Probability of any toxic gas and aerosol releases to the atmosphere due to any emergency situations shall be analyzed. Assessment methods and values of the toxicity parameters for these emergency situations shall be described.

Probability of gas and aerosol ingress to the rooms shall be analyzed, and safety of the workers shall be assessed.

3.5.2.4. Radiation impacts

In case any damages of buildings and (or) structures containing radioactive substances are possible due to emergency situations at the NPP site radiation intensity as well as parameters of radionuclide dispersal processes in the atmosphere, surface and ground water shall be defined.

3.5.2.5. Fire load

It should be explained in brief how the fire load is formed in case of any fires in fire-hazardous rooms with sodium, oils, in cable and other rooms, and what load combinations it can contribute to. It should be demonstrated for what structures safety factors shall be substantiated in consideration of the fire loads. Results of review and analysis shall be presented in the relevant sections of the NPP SAR.

3.6. Design load combinations for the NPP power unit buildings and structures

General approaches to assignment of load combinations caused by external impacts of natural and human-induced origin, internal impacts due to emergency situations at the NPP site and inside the main building, the impacts occurring under normal operation conditions (particularly in transient modes) shall be described.

It should be demonstrated that the combinations of loads on the buildings and structures selected for consideration are assumed in accordance with the RD requirements. Combination of loads on the NPP power unit buildings and structures shall be described.

All types of loads on the buildings and structures shall be presented in the tabular format.

It should be specified in what structures and buildings and for what elevations the floor accelerograms and response spectra shall be taken for further analysis of resistance of the equipment, pipelines and other systems and components to external impacts.

3.7. Protection of the NPP territory against geological hazards

Description and substantiation of the measures for protection of the territory against geological hazards to be arranged with due regard for the RD requirements shall be provided.

The lists of design materials containing the information on the engineering measures aimed to eliminate, mitigate the consequence and observe the development of geological hazards described in Section 2 shall be presented. The general map of the design arrangements for protection of the NPP territory including the flooding protection arrangements (flow control, removal of surface and ground water), installation of mud barriers and dams, strengthening of sliding and undercut slopes, etc. as well as adequacy evidence for the protective measures and characteristics of external impacts changed due to protection arrangement shall be provided.

3.8. Flooding protection

Measures for protection of safety-related buildings, structures, components and systems against flooding shall be described. In this case:

- structures where any safety-related equipment is installed shall be described, any apertures and passages located below the design flooding level (if any) shall be indicated;

- systems and components subject to protection against flooding shall be defined;

- methods for determination of static and dynamic impact of the design flooding or groundwater on safety-related buildings and structures shall be described;

- means for protection of the equipment against flooding (for example water removal pump systems, stop log gates, water-tight doors and drainage systems) shall be described;

- protection against ingress of water to the structures, elimination of water leakages and impact of wind waves (including splashing) shall be described. Individual chambers, compartments and cells where safety-related equipment is installed that serve as natural barriers to prevent their potential flooding shall be indicated on the layout schemes;

- methods of flooding protection with calculation of the time necessary to provide protection shall be presented.

3.9. Justification methods and criteria for ensuring mechanical strength of the NPP power unit buildings and structures

All applied justification methods and criteria for ensuring mechanical strength of the NPP power unit buildings and structures shall be described in order to confirm their applicability for calculations of the buildings and structures in accordance with the classification (Subsection 3.2, Section 3) and types of impacts.

3.9.1. Safety-related buildings, facilities, civil structures and foundations

It is necessary to describe methods for design substantiation of the mechanical strength of safety-related buidlings, facilities, civil structures and foundations in relation to:

- external impacts specified in Section 2;

- impacts caused by emergency situations at the NPP site external in relation to the reactor building (Subsection 3.5, Section 3).

Methods considering the peculiarities of the buildings, facilities and their components (leak-tight rooms, foundations, civil structures) shall be described, or references to Section 3 where they are specified in detail shall be given.

The mechanical strength criteria (strength, leak-tightness, fire resistance, seismic resistance, etc.) shall be formulated. Compliance with this requirement shall be demonstrated in the relevant paragraphs of Section 3.

It shall be specified that the applied methods for justification of mechanical strength of buildings, facilities, building structures and foundations under external impacts comply with state-of-the-art in science and technology. In case any simplified methods are applied their acceptability shall be confirmed.

3.9.2. Hydrotechnical and geotechnical structures, units and channels

Requirements for hydrotechnical and geotechnical structures, units and channels shall be specified in order to ensure their mechanical strength under static and dynamic impacts mentioned in Section 2 with regard to each impact type and their possible combinations.

Methods and methodologies used to analyze mechanical strength in relation to each impact type and the selected load combinations shall be presented.

3.9.3. Software used

The list of software tools used to justify mechanical strength of buildings and structures particularly under external impacts shall be provided.

The following information shall be presented for each program:

- brief description of the program purpose;

- the calculation method implemented in the program;

- basic restrictions and assumptions;

- information on the program validation in the state regulatory authority for safe atomic energy use;

- results of the program verification by analytical and experimental methods (in case no program validation has been performed).

3.9.4. Methods for testing and full-scale investigations of the buildings, facilities and structures

In case any modelling methods of testing are used along with calculation methods to analyze mechanical strength of the buildings, structures and facilities the following information shall be provided:

- criteria and applied modelling methods;

- description of the methods applied to test the models of buildings, facilities and structures;

- description of test stands and testing equipment;

- techniques and methods for determination of dynamic characteristics for the buildings, facilities and structures;

- methods for setting of impacts and determination of the load level;

- mechanical strength determination criteria for the buildings, facilities and structures;

- methods for assessment of test errors and adequacy of the obtained results.

The following information shall be presented for full-scale investigations of the buildings, facilities and structures:

- methods and programs for full-scale investigations of the buildings, facilities and structures;

- methods of impact setting;

- criteria for selection of the response record points;

- techniques and methods for determination of dynamic characteristics for the buildings, facilities and structures;

- mechanical strength determination criteria for the buildings, facilities and structures subsequent to the testing results;

- equipment and instrumentation;

- methods for assessment of investigation errors and reliability of the obtained results.

3.9.5. Mechanical strength criteria for the buildings, facilities and structures

The list of safety-related buildings, facilities and structures shall be provided, and their limit states shall be established. The limit states shall be considered as the operability criterion. These data shall be presented in the table according to the approximate format given in Table 3.2.

Table 3.2

No. Buildings, facilities and structures

Limit states

Parameters Numeric value

Other parameters

3.10. Determination of loads transferred to the equipment, pipelines, systems and components via the civil structures under dynamic impacts of natural and human-induced origin

Methods applied to determine loads on the NPP power unit systems and components for more detailed analysis of their mechanical strength under external and internal dynamic impacts shall be described.

3.10.1. Input data for dynamic calculations

Approach to the arrangement of the NPP power unit facilities subject to dynamic analysis and the possibility to divide the facilities into independent sub-systems shall be analyzed. The following information shall be presented for each facility:

1. Basic characteristics of the facility:

- physical dimensions;

- total weight;

- weight distribution by sub-systems.

2. Arrangement of the foundation beds (facilities with common foundation beds shall be specified).

3. Mutual arrangement of individual foundations in order to consider their impact on the stress state of the bases.

3.10.1.1. Accelerograms (seismic analysis)

The set of applied DBE and SSE accelerograms for horizontal and vertical ground motion shall be

provided.

The main parameters (maximum acceleration, the basic frequency, effective accelerogram duration, rise and fall time of the accelerogram amplitude) shall be determined.

All design accelerograms selected from the available records of past earthquakes or obtained through the use of well-known synthesis methods on the basis of response spectra shall be substantiated. The methods used to select accelerograms for calculations shall be specified, and their applicability shall be substantiated.

The maximum residual displacement shall be specified for accelerograms.

The corresponding response spectra for various damping values used in the design of structures, systems and components shall be provided for the accelerograms selected for impact analysis. Frequency ranges for calculation of the spectral values shall be specified.

Comparison of the response spectra obtained in the free field on the ground surface and at the foundation level of the facilities referred to the first seismic category with the design spectra shall be performed for each damping value used in the design of facilities. It should be demonstrated that the design accelerograms are compatible with the design response spectra (see par. 3.10.1.2).

The method for application of the selected set of accelerograms for systems and components shall be described.

3.10.1.2. Response spectra (seismic analysis)

The response spectra used to substantiate seismic resistance of the buildings, facilities and structures at the locations of the NPP power unit buildings referred to the first seismic category on the ground surface and at the foundation level of the facilities shall be provided.

Response spectra shall be provided for various damping coefficients of horizontal and vertical ground motions.

Sources used as the basis for selection of the design response spectra shall be specified and the selection shall be justified.

The method for application of the design response spectra in the dynamic analysis shall be described.

3.10.1.3. Soil modelling

Soils at the base of each facility referred to the first seismic category shall be described including the following information: the foundation penetration depth, the basic physical dimensions of the foundations, thickness of soil above the rock bottom, characteristics of the soil strata, the total weight of the structure. The mathematic model of the soil used in further dynamic calculations shall be described. In case the model of a multi-layer base with the underlying half-space is used the following soil characteristics shall be specified for each layer: shear wave speed, specific gravity, thickness of layers, Poisson's ratio and damping.

The presented information shall have the scope required to assess interaction between the soil and the structure by the finite element method or by the equivalent resilience method.

3.10.1.4. Damping coefficients

Information on the damping coefficients shall be provided, and the applied damping coefficients for the soils as well as for the facilities referred to the first seismic category and their internal structures shall be substantiated. Techniques and methods for determination of the damping coefficients shall be

described, or the sources used for selection of these coefficients shall be specified.

3.10.2. Methods for dynamic behavior analysis of the structures.

Methods used to analyze dynamic behavior of the buildings and structures referred to the first seismic category shall be described. Besides, specific information listed in the subsections below shall be included.

3.10.2.1. Analysis methods

Standard mathematical models used to calculate the vibration parameters for the facilities and structures referred to the first seismic category shall be described with indication of the characteristic peculiarities used in the course of modelling. Selection of any model shall be substantiated.

The method used in the seismic analysis for determination of the maximum relative shift of the supports shall be specified.

If the response spectrum analysis method was used the criteria for selection of the number of natural modes sufficient for analysis shall be specified.

Any other important factors to be taken into consideration for seismic analysis (for example, hydrodynamic, non-linear, micro-deformational effects and characteristics of interaction with the main structures) shall be indicated.

3.10.2.2. Modelling methods

Criteria and methods applied in the calculation schemes within the framework of the selected model shall be provided.

Calculation schemes used for determination of the dynamic characteristics shall be described for all facilities referred to the first seismic category. Selection of particular calculation schemes shall be substantiated. In case different models or calculation schemes of the facilities are used in calculations for various external impacts their descriptions shall be provided.

Dynamic characteristics calculation results obtained for different models (schemes) of the facility shall be compared.

The basic dynamic characteristics shall be specified for each facility. In case the response spectrum analysis is used in the calculations the following information shall be provided for each mode: frequency, modal mass, modal damping. Uncertainty of the results introduced by reduction of the number of modes used for calculations shall be assessed.

Dynamic characteristics of the facilities shall be provided for the schemes with and without regard to the soil flexibility influence. Impact of the soil-to-structure interaction effects shall be assessed.

Peculiarities of the facility modelling for separate calculation of their dynamic characteristics per each dynamic impact shall be presented.

Criteria and input data required to determine the necessity for investigation of any unit as a part of the analyzed system or as an independent sub-system shall be specified.

3.10.2.3. Soil-to-structure interaction

Methods for calculation of the soil-to-structure interaction shall be described, and their application shall be substantiated.

In case the equivalent resilience method is applied the methods to obtain the parameters used in the analysis shall be specified.

The methods used in the analysis to take into account physical and mechanical properties of soils, attitude of strata and changes of the soil properties shall be also described. Applicability of the equivalent resilience method for the conditions of the particular site shall be justified.

Any other methods for analysis of the soil-to-structure interaction or substantiation of such analysis rejection shall be specified. Criteria and methods used to take into account the impact of the adjacent structures on the response of the structure under consideration shall be provided in the analysis of the soil-to-structure interaction.

3.10.2.4. Interaction of structures

Approaches to consideration of the interaction between the structures located on the common foundation or on separate foundations shall be described. Criteria for consideration of joint seismic vibrations of the structures or their parts, particularly the ones not referred to the first seismic category, in the seismic analysis for the structures referred to the first seismic category or any parts thereof shall be provided.

3.10.2.5. Earthquake impact in three mutually perpendicular directions

The way to consider the earthquake impact in three mutually perpendicular directions in order to determine seismic response of the structures, systems and components and its compliance with the RD requirements shall be described.

In case the static method is used for the vertical direction in seismic analysis of the structures, systems and components, and the dynamic analysis method or the response spectrum method is used for the horizontal directions the applicability of this approach shall be substantiated.

3.10.2.6. Method used to consider the torsional impact of earthquakes

In case the static method or any other approximation method is used to calculate the structures referred to the first seismic category instead of the joint dynamic analysis of these structures under vertical, horizontal and torsional impacts the possibility to apply these methods shall be justified. The method used to consider the torsional impact in the seismic analysis of the structures referred to the first seismic category shall be described.

3.10.2.7. Combination of natural vibration modes

In case the response spectrum method is applied information on the procedure used to summarize the relevant vibration modes and to determine force factors and displacement factors (shifts, moments, stresses, deflections and accelerations) shall be provided.

3.10.2.8. Basic results of dynamic calculations


- dynamic characteristics of the structures obtained for the schemes with due regard for the interaction between the soil and structures with fixed base;

- impact of the consideration of the soil-to-structure interaction effects on the basic dynamic characteristics;

- vibration parameters for the facilities and structures;

- dependence of maximum displacements from the elevation;

- dependence of maximum accelerations from the elevation.

3.10.2.9. Floor accelerograms and response spectra

Methodologies used to obtain floor accelerograms and response spectra with due regard for three soil vibration components shall be described. In case the response spectrum method is used to determine floor response spectra conservatism of this method in relation to the over-time direct integration method shall be substantiated. Methods used to obtain design floor response spectra (criteria for development of the envelope curves, their smoothing, expansion of peaks, etc.) shall be described.

Methods for determination of the design floor accelerograms corresponding to the design response spectra shall be specified.

Criteria for selection of loads obtained under various external impacts for their further use in the mechanical strength analysis for the NPP systems and components shall be specified and substantiated.

The methods used to consider the impact of uncertainties in structural and physical and mechanical soil properties on the soil-to-structure interaction, floor response spectra or floor accelerograms shall be described.

3.10.2.10. Seismic isolation of the structures and other arrangements for correction of vibration parameters

Seismic isolation of the structures used to reduce dynamic, seismic, shock and vibration impacts on the systems and components located in them, substantiations of its reliability as well as rules for acceptance into operation and control in the course of operation shall be described.

Conclusions on impracticality of seismic isolation based on the soil-to-structure interaction analysis shall be given for any structures referred to the first category where no engineering seismic isolation features are installed.

Techniques for protection of all structures referred to the first category against seismic and other dynamic impacts, the scope of compensatory measures shall be described, and efficiency of the RB seismic isolation shall be assessed.

3.10.3. Dynamic loads from the impacts of non-seismic origin

Methods used to determine time dependence of the resulting loads for the dynamic loads of non-seismic origin (aircraft crash, blast wave, etc.) selected for consideration in the design shall be described.

Methods used to determine the load at the impact point (methods to solve the contact problem with collision of two bodies) shall be described for the impact of "aircraft crash" type.

In case the non-linear interaction method is applied it is necessary:

- to substantiate selection of this method;

- to specify criteria and justification for selection of the load application directions and points.

For the impact of "blast wave" type it is necessary:

- to describe the methods used to determine the load;

- to specify the criteria for selection of the load application directions and points.

3.11. Buildings, facilities, civil structures, bases and foundations

Constructive solutions for the buildings, facilities, civil structures, bases and foundations shall be described, substantiation of their strength, leak-tightness, fire resistance and mechanical strength under external impacts shall be specified in brief, and the arrangement for strengthening of the bases and foundations for the safety-related buildings, facilities and structures shall be also listed and substantiated.

The complete list of documents containing substantiation of the constructive solutions for buildings, facilities, civil structures, bases and foundations, seismic isolation as well as description of the testing and operability control programs for the structures shall be provided. Strength of the safety-related buildings, facilities and civil structures shall be substantiated.

3.11.1. Main building

3.11.1.1. Description of the buildings, facilities and civil structures of the main building

Approach to arrangement of the facilities constituting the main building shall be analyzed. The following information shall be presented for each facility:

1. Basic characteristics of the facility:

- physical dimensions;

- volume;

- total weight;

- weight distribution by sub-systems.

2. Arrangement of the foundation beds (facilities with common foundation beds shall be specified).

3. Mutual arrangement of individual foundations in order to consider their impact on the stress state of the bases.

4. Expansion, settlement and seismic joints in the structures, between the posterns and passages.

Information on the dimensions of the facilities, degree of prefabrication for the structures, applied materials (types, classes and grades of concrete and reinforcement bars) in the structural elements and their design characteristics shall be provided for all components of the facilities.

The information shall be presented for all safety-related structures of the main building including the reactor building.

3.11.1.2. Summary table of impacts on the buildings and civil structures of the main building and their combinations

The summary table of impacts and their combinations considered for the main building facilities shall be provided.

3.11.1.3. Stability assurance for bases and foundations of the structures

Stability of the bases and foundations for the structures shall be substantiated, and the information on any engineering arrangements aimed to ensure stability of the bases and foundations

shall be provided.

Measures aimed to prevent any unacceptable deformations of the bases due to possible groundwater level increase, under the impact of any static and dynamic lods, in case of soil liquefaction (drainage, soil stabilization, etc.) and due to any other geological processes and phenomena classified as hazardous ones shall be described.

Information on the calculations of interaction between the bearing surface of the foundations and soils shall be provided.

Impact of any other mutually arranged foundations and structures on the stress state of the base under consideration shall be assessed.

The following information shall be provided for each foundation:

- basic reinforcement, floor lining with the anchorage system;

- the anchorage system of the internal structures for fastening to the foundation bed (and also options of anchorage through the lining);

- mechanics of the foundation shear strain under horizontal loads (for example, seismic impacts), the technique for transfer of horizontal loads to the shock absorbers;

- the layout of shock absorbers;

- assessment of the foundation capability to take up shear forces in presence of waterproofing.

3.11.1.4. Assessment of interaction between structures and bases

Design limits of the parameters characterizing stability of each structure and its foundation including differential settlement and strength margins against tripping and creepage shall be specified.

Deformation and load bearing capacity analysis results with description of the calculation method for settlement, tilt and stability (estimated settlement for the construction period and the operation period with due regard for load increase over time) shall be provided.

3.11.1.5. Foundation examination and monitoring

In case continuous examinations and monitoring of the foundations are required due to geological conditions the program of the above-mentioned examination and monitoring and engineering features for the foundation state control shall be described.

Requirements for stress state monitoring of the base soils and estimations of the foundation settlement shall be specified.

Information on the program for monitoring of the foundation settlement and the structure tilt within the NPP construction and operation period as well as the applied monitoring hardware shall be presented.

3.11.1.6. Strength and stability assurance

Results of strength, leak-tightness, fire resistance and mechanical strength under external and internal impacts shall be provided for the structures of the main building.

3.11.1.6.1. Civil structures of the reactor building

The list of safety-related civil structures of the reactor building, loads and load combinations and

limit states shall be provided.

The most important civil structures of the reactor building shall include at least:

- external envelope structures;

- the reactor support system;

- the RCP support system;

- the reactor cavity;

- the ceiling structures;

- the overhead crane support structures;

- waterproofing of the rooms with SS electric equipment and the rooms with any equipment containing liquid sodium.

The above-mentioned list may be supplemented and expanded in each particular project.

Layout and constructive solutions of the reactor building including the drawings of the internal structures shall be described. References to the materials containing strength and stability substantiation for the internal structures shall be given. Calculation schemes for the internal civil structures with substantiation of the applied assumptions and conclusions subsequent to the results of dynamic load calculations for the internal civil structures of the reactor building, as well as information on the materials, reinforcement, loads on the equipment installed on these structures shall be provided.

The list of all rooms where a fire is likely to break out with indication of the potential fire hazard causes shall be presented.

Substantiated information on compliance with the requirements for fire resistance of the internal structures shall be provided.

The in-service inspection program for the internal civil structures of the reactor building shall be presented. In case any new construction methods are used the scope of testing and in-service monitoring shall be defined.

3.11.1.6.2. Information on concrete, its components and steel reinforcement

Information on concrete, its components (cement, gravel, sand, water) and steel reinforcement shall be provided.

Selection of the materials shall be substantiated with due regard for the normal operation conditions, accidents, compliance with the requirements for compatibility of the structural materials with the coolant, compatibility of the structural materials with heat insulation materials and compatibility of the latter with the coolant.

The applied validated calculation programs shall be specified.

Conclusions on strength, stress-strain behavior, crack resistance of individual structures and the entire facility shall be made based on the comparison of the obtained calculation results for the adopted models with the standard criteria.

Safety factors for stresses and forces in the reinforcement steel and concrete, deformations and crack resistance shall be determined in order to assess efficiency of the constructive solutions subsequent to the calculation of the assumed load combinations.

Construction methods shall be described, and information on the applied structural materials and expected changes of their properties in the course of operation shall be provided.

In case any new construction methods are proposed for use they shall be described.

References to the developed quality control programs for the materials and works shall be given.

Information enabling to determine compliance of the adopted quality control programs to the RD requirements shall be presented.

Quality control programs for the materials shall be described including tests aimed to determine physical and mechanical properties of concrete, reinforcement steel, fasteners, lining sheets and anchorages. Control methods for the pre-stressing system (if any) shall be presented.

Requirements for tests and inspections during operation of the structures shall be specified.

The final objective of testing and the adopted criteria for assessment of the results shall be specified. In case any new construction methods that have never been applied before are used the scope of additional tests and in-service inspections shall be defined indicating the degree of compliance of these tests with the requirements of the in-service inspection programs. Information on inclusion of the in-service inspection programs to the technical specifications shall be provided.

3.11.2. Other NPP power unit buildings and facilities not included into the main building

Descriptions and substantiations of strength, leak-tightness, fire resistance and mechanical strength under external impacts shall be provided for any other safety-related buildings and structures, their foundations and internal civil structures.

They include:

- the turbine hall building;

- the SDGS building;

- the service water pump station for essential consumers of the NPP;

- the spray pond for water supply of the essential consumers of the NPP power unit (if any);

- the special building;

- water intake structures, tunnels, channels;

- the underground diesel fuel storage facility;

- the building of power supply sources referred to the first category (a storage battery, inverters, uninterruptible power supply units);

- the building of the beyond design basis accident management center (in case of a separate building);

- PPS buildings and facilities of the NPP power unit;

- RW storage facilities;

- buildings and facilities of the SS fire-fighting pump station;

- the FFS building.

The above-mentioned list shall be considered as approximate and may be supplemented and adjusted for each NPP power unit. Detailed information shall be provided for each of the above-mentioned buildings and facilities. Information shall be presented in accordance with the most acceptable structure complying with the peculiarities of the buildings and facilities; it shall also contain the conclusions on stability of the bases and foundations.

Classification of the FFS buildings, facilities and civil structures in accordance with the requirements of the Construction Design Standards for NPPs with Various Types of Reactors and Design Standards for Seismic-Resistant Nuclear Power Stations as well as information confirming their compliance with the criteria established in these RDs shall be provided.

In case there are any dams, weirs and other structures posing hazard to the NPP near the NPP mechanical strength under external impacts shall be assessed for each structure, and the arrangement for reinforcement of the bases shall be described.

Conclusion on strength and stability of all buildings, facilities and civil structures shall be made on the basis of the calculation and analysis results.

3.11.3. Diagnostics of civil structures

Description of the diagnostic system for the facilities and civil structures including the system for monitoring of tilting, settlement, stress-strain behavior, vibrations and foundation conditions shall be provided. Particular facilities and civil structures subject to mandatory diagnostics in order to ensure safety of the NPP power unit shall be specified. Information on equipment of the buildings and facilities with reference marks, the systems for monitoring of tilt, settlement and vibrations of buildings and structures, the foundation conditions as well as their stress-strain behavior shall be provided.

Information on the observation program shall be presented for the above-mentioned observations.

Information on settlement and tilts of the buildings and structures, stresses in the structures and foundations recorded after installation of the equipment prior to the NF loading based on the actual state of the facilities subsequent to their testing and according to the observation data shall be provided.

3.11.4. Investigation program and action plans for inspections of the essential NPP buildings and facilities

The list of planned investigations and inspections of the foundations, buildings, facilities, civil structures, soil conditions, groundwater, monitoring of the general state of the facilities and control of radiation leakages in the wells shall be provided.

These investigations and inspections shall be described in brief.

3.12. Substantiation methods for strength and operability of the NPP power unit equipment, pipelines, systems and components with due regard for the loads caused by natural and human induced impacts and transferred via the civil structures, buildings and facilities

Information containing the basic calculations for substantiation of strength and operability for the NPP power unit equipment, pipelines and components, determination of the capability of the mechanical, instrumentation and electrical systems to perform their functions under combined impact of external conditions, internal emergency impacts and normal operation impacts shall be presented.

3.12.1. Consideration of external conditions in analysis of the mechanical, electrical and instrumentation equipment

Information on any external conditions taken into account for design of the mechanical, electrical and instrumentation equipment shall be provided.

3.12.1.1. Testing methods for systems and components

Methodologies, stands, testing equipment used to substantiate mechanical strength of the NPP systems and components shall be described. It is recommended to present the information in the following order.

1. Operability testing and investigation

Tests and studies performed or to be performed for each component in order to check its operability under the combination of such impacts as temperature, pressure, humidity, chemical composition and radiation shall be described. The particular values of the impact parameters (temperature, pressure, etc.) shall be specified.

2. Vibration testing methods

Criteria and methods of vibration testing and dynamic analysis applied to confirm structural and functional integrity of the systems, pipelines, mechanical equipment and the reactor internals subjected to vibration loads including any loads caused by the coolant flow shall be described.

3. Check tests of the equipment for operability under external impacts

Information concerning only the impacts of non-seismic nature shall be presented.

3.12.1.2. Software used

The list of software tools used to substantiate mechanical strength of the NPP power unit equipment, pipelines, systems and components under external impacts shall be given. The following information shall be presented for each program:

- brief description of the program purpose;

- the calculation method implemented in the program;

- basic limitations and assumptions for the program;

- information on the program validation.

3.12.2. Mechanical systems, equipment components and pipelines

3.12.2.1. Strength and stability analysis

Methods for analysis of strength and stability of the mechanical systems, equipment components and pipelines shall be described. Complete input data for strength analysis in the course of the NPP power unit operation and in case of any anticipated operational occurrences shall be provided (or the reference to the section contaning such information shall be given).

The list of computational programs used for static and dynamic analyses of the structural and functional integrity, strength and stability of all safety-related systems, assemblies, equipment and support structures shall be presented.

Methods used to assess stresses under emergency conditions as well as experimental methods for stress analysis, particularly in case when these methods are applied instead of calculation methods shall be described.

In case any creep deformations can occur in the equipment under emergency conditions the methods used to determine deformations and stresses as well as the adopted criteria shall be described.

3.12.2.2. Dynamic testing and analysis of mechanical systems, equipment components and pipelines

Criteria and methods for testing and dynamic analysis applied to confirm structural and functional integrity of the mechanical systems, equipment components, pipelines and the nuclear reactor internals subjected to vibration loads, including the loads caused by the coolant flow and seismic impacts shall be provided.

3.12.2.2.1. Pre-operational, vibration and dynamic testing of the pipelines

Information on availability of the testing programs shall be provided.

3.12.2.2.2. Seismic resistance testing and checks for the safety-related mechanical systems, equipment and components

Information on seismic resistance tests containing the following data for all types of safety-related mechanical systems, equipment and components shall be provided:

- description of the seismic resistance criteria, testing methods, the main parameters of testing regimes, the way to consider the impact of the equipment elevation on the parameters of the selected testing regimes;

- substantiation of the testing program adequacy for determination of the seismic characteristics of the equipment.

Techniques and methods for analysis and testing of the supports for mechanical equipment shall be described.

Conclusions on the seismic resistance of mechanical systems, components and equipment shall be provided.

3.12.2.3. Calculation analysis, pre-operational testing of the reactor internals for vibration caused by the coolant circulation

The analysis method used to investigate the behavior of the structural elements located inside the nuclear reactor pressure vessel in normal and transient modes of the coolant circulation shall be described.

The analysis shall be applied to determine the force loads affecting the reactor internals on the coolant side and to predict vibration characteristics of the reactor internals.

Information on selection of the mathematical model and acceptance criteria for the structures as well as the information demonstrating the peculiarities in location of the points where the vibration characteristics are calculated shall be provided.

Information on the pre-operational testing of the reactor internals for vibration loads caused by the coolant circulation within the functional check program in the course of the commissioning works shall be provided.

3.12.3. Electrical equipment

The methods used to substantiate operability of the elctrical equipment shall be described, and information confirming compliance with the requirements of the regulatory document "Seismic Resistance of the NPP Automation Means. Technical Requirements and Testing Methods" shall be provided.

3.12.3.1. Criteria for the electrical equipment operability check under dynamic loads

Seismic resistance checking criteria including the criteria for selection of the testing methods, methods for setting of the input parameters of vibration, types of loads used to check operability of the electrical equipment as well as their values with due regard for the equipment location at the NPP power unit shall be described.

3.12.3.2. Techniques and methods for the equipment mechanical strength and operability checking under loads

Techniques and methods used to check seismic resistance of the electrical equipment referred to the first seismic category (calculation and testing) shall be described.

3.12.3.3. Techniques and methods for stability analysis of the support structures

Techniques and methods for the calculation analysis or testing of the support structures for electrical equipment referred to the first seismic category for stability under dynamic loads shall be described.

3.12.4. Thermomechanical equipment

Criteria used in the course of testing or analytical studies in order to justify operability of the thermomechanical equipment shall be defined. Brief description of the testing programs and calculation methods and the applied combinations of loads shall be presented.

Basic conclusions on the results of strength analysis and operability assessment for the thermomechanical equipment shall be provided.

Techniques and methods for checking the stability of the support structures for the thermomechanical equipment under the selected combinations of effective loads including external impacts shall be described.

Calculation methods used to substantiate strength and operability of the steam generator with due regard for the loads caused by external impacts shall be described. The applied calculation schemes shall be provided, and their conservatism shall be substantiated. Combination of loads used in the calculations, the methods and conclusions subsequent to the calculation results obtained with due regard for the loads caused by the jet impact in case of a pipeline breakage, reaction thrust, external impacts, emergency loads as well as the applied strength criteria shall be described.

The methods used for calculation and analysis of the SG supports under the selected load combinations shall be presented.

3.12.5. Diesel generators

The rooms of the diesel generators shall be described including the general drawings with the required sections enabling to determine mutual arrangements of the diesel generators and the adjacent structures. Calculation schemes and combinations of loads used in the calculations, description of the calculation methods with due regard for the assumptions, the mechanisms of load transfer from the

foundations to the diesel generators under external impacts shall be provided. The computational programs used shall be specified.

3.12.6. Instrumentation and control hardware

The range of the instrumentation and control hardware referred to the first seismic category and conditions for their location and fastening to the structures shall be described. Criteria for checking of seismic resistance and stability against external impacts shall be specified. The loads used to check seismic resistance and stability against external impacts with due regard for the location, techniques and methods applied to check the instrumentation and control hardware for mechanical strength under external impacts shall be described.

Techniques and methods applied to check the supports of the structures where the instrumentation and control hardware is installed for mechanical strength under external impacts shall be described.

Conclusions shall demonstrate performance of the relevant safety functions by these instruments and equipment even after external impacts considered in the design.

3.12.7. Ventilation equipment and air ducts, equipment of the filtration systems

Strength and stability of the ventilation equipment and air ducts as well as the equipment of the filtration systems under the loads defined in Subsection 3.5 shall be substantiated.

The range of the equipment, the list of safety-related air ducts and filtration systems shall be provided.

The sources containing complete analysis of strength and stability under the impacts of internal origin and natural and human-induced external impacts shall be specified. Conclusions on strength and stability with indication of the following shall be provided:

- design loads and their combinations;

- calculation and analysis methods, modelling;

- testing methods, test stands and testing equipment;

- criteria for stability and strength of the ventilation equipment, air ducts, filtration systems;

- techniques for fastening on the structures, strength of the support assemblies, explanatory diagrams and drawings.

3.12.8. Handling equipment

The range of the handling equipment shall be described, and the places for its location shall be specified. The fastening techniques shall be described, and the explanatory diagrams and drawings shall be provided.

Strength, resistance and stability of the handling equipment shall be substantiated with due regard for the complete range of external and internal impacts defined in Subsection 3.5. Evidence of acceptability of the methods selected for substantiation and reliability of the results as well as information on the strength, resistance and stability criteria and the testing programs shall be provided.

4. REQUIREMENTS FOR THE SECTION "REACTOR AND

THE PRIMARY CIRCUIT SYSTEMS"

Information and analysis results required to substantiate safe operation of the reactor and the primary circuit systems within the design service life of the reactor plant under normal operation conditions and in case of any operational occurrences (including accidents) as well as the information required to perform the analysis in order to obtain the results specified in Section 15 shall be provided.

Information and analysis presented in this section shall be based on the design materials for the RP, the nuclear core, the reactor internals and any other safety-related systems, the results of S&RW and R&DW.

4.1. Purpose of the reactor and the primary circuit systems

4.1.1. Purpose and functions

Purpose and functions of the reactor and the primary circuit systems shall be specified.

Information on the regulatory basis of the RP design shall be provided in the form of the list included into the appendix to Section 4.

It should be noted that the reactor and the primary circuit systems are designed as safety-related normal operation systems consisting of the components referred to the first, the second and the third safety class (the particular class shall be specified in the description of the relevant equipment) but containing a safety system arranged in the form of a safety containment and intended for the coolant confinement in case of any leaks from the reactor pressure vessel (except for the cover).

All equipment located in the reactor pressure vessel shall be referred to the first seismic category and shall be designed for the seismicity corresponding to the SSE.

4.1.2. Design basis


- design characteristics of heat energy generation;

- the nuclear fuel used;

- characteristics of the design;

- the NF use mode;

- the NF burn-up;

- duration of the RP operation within a year;

- design service life of the RP;

- maintainability and restorability;

- the primary circuit systems.

The RD provisions (GSR, NPP RF NSR, etc.) should not be provided as they include mandatory safety requirements and not design basis.

4.2. Reactor design

4.2.1. Description of the reactor

Description of the reactor with the reference to the relevant design documents shall be provided.

Information on installation of the reactor in the cavity and brief information on the building where the reactor is located, protection of the reactor building against external natural and human-induced impacts (see Section 2) and events at the NPP site external in relation to the reactor building shall be provided.

Coordinates of the reactor shall be specified.

Orientation of the reactor in relation to the NPP building, mutual arrangement and interaction of the described equipment and systems and their mutual impacts shall be understandable from the description.

The description shall include the list of the constituent parts - the reactor systems (components) performing independent functions. The list shall include:

- the nuclear core;

- the reactor shutdown system - the EP (CPS) control rods;

- the passive emergency system;

- the CPS (actuators and the drive);

- the reactor pressure vessel including the reactor internals;

- the safety containment of the reactor;

- rotating plugs;

- the in-core fuel assembly handling equipment (system);

- the sodium purification system;

- the reactor cover gas makeup system (within the primary circuit boundaries);

- other systems and components (for example special-purpose channels);

- the primary circuit systems located in the reactor pressure vessel (the pressure header, the primary and secondary circuit IHE, etc.).

4.2.1.1. Nuclear core

4.2.1.1.1. Purpose and design basis

The purpose and design basis of the nuclear core and its assemblies shall be described, their groups in accordance with the safety and seismic resistance classification shall be specified, the list of RDs defining the design criteria and safety principles, the main requirements for arrangement of the nuclear core and design of its assemblies shall be provided.

In case of any refurbishment of the reactor nuclear core related for example to usage of new fuel types the design materials for such refurbishment and the additional safety analysis materials shall be provided.

4.2.1.1.2. Description of the nuclear core configuration

The nuclear core configuration and structure of its assemblies shall be described, their general outlines showing mutual arrangement, basic dimensions, fastening methods and orientation in relation to the reactor axes, diagrams of the coolant distribution for the core assemblies shall be provided.

Fuel load patterns shall be provided for the first loading of the core, intermediate loadings and steady operation mode of the reactor, and information on the NF quantity shall be presented. Reference to the relevant drawing from the technical project list for the nuclear core and fuel assemblies shall be given for each figure presented.

Description of the nuclear core and its assemblies shall be accompanied with the list of their basic technical characteristics.

4.2.1.1.3. Materials, nuclear fuel, coolant

Selection of the materials for the nuclear core assemblies shall be substantiated, the nuclear fuel and the coolant shall be described; in this case the following information shall be provided:

1. For structural materials:

- mechanical, thermal and physical properties depending on the exposure dose and temperature (yield point and breaking point, residual plasticity, thermal conductivity, thermal capacity);

- strength and thermal creepage depending on the exposure dose, temperature, load and exposure time;

- corrosive interaction with fission products and the coolant depending on the NF burn-up, temperature and the NF irradiation period;

- cyclic strength depending on the exposure dose, temperature, load and number of cycles.

2. For nuclear fuel:

- chemical composition, enrichment, density, loading, irregularities in distribution of density and fissionable isotopes, methods for their control, validation of the control methods;

- the NF creep flow and swelling depending on temperature, exposure dose and load;

- mechanical, thermal and physical properties depending on the burn-up level, temperature, content of fissionable isotopes (melting temperature, thermal capacity, thermal conductivity, heat expansion, breaking point);

- compatibility with the cladding material, mass transfer depending on burn-up, temperature and time;

- the posiibility and feasibility of SNF processing (brief information).

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the results of investigations for qualification of such fuel, for example by its irradiation in research reactors or irradiation of pilot assemblies with the new fuel type in the operating reactors, etc., as well as estimates of the permissible burn-up depth shall be additionally provided.

3. For absorbing materials:

- chemical composition, dimensions, enrichment, density, control methods, validation of the control methods;

- compatibility with the cladding materials;

- behavior in case of any accidents;

- behavior under irradiation and changes of properties.

4. For the coolant:

- thermal and physical properties;

- permissible impurities;

- specific properties and peculiarities providing for its usage as the coolant for the fast reactor.

4.2.1.2. Reactor cavity

Description of the reactor cavity shall be presented.

4.2.2. Control and monitoring

The list of the controlled parameters for the nuclear core and fuel assemblies, frequency of control, the range of parameter measurements, permissible measurement errors, configuration and location of sensors shall be provided and substantiated.

Information on the nuclear core state monitoring and the RP power control shall be presented:

- protections and interlocks, controllers, diagnostic systems, automated control programs;

- for reactivity control - the absorbing rod system - the EP (CPS) and PEP control devices representing independent systems;

- for neutron flux measurement - the neutron flux control system which is a normal operation system but is arranged in accordance with the requirements for CSSs due to its importance for safety;

- for any position changes of the control devices - the drive control system (a part of the CPS); this system shall be described in par. 4.2.5 of Section 4 (may be described in Section 7);

- the in-core monitoring system;

- the state diagnostics system for the safety barrier - the fuel element claddings (in case such system is provided);

- the RP power control and limitation system;

- the system for generation of the preventive protection and interlock commands (in Sections 7 or 12, in the subsection related to CSSs if these commands are generated in the EP CSS);

- the system for generation of the RP emergency shutdown commands - the EP CSS (described in Section 12).

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type applicability of the existing metrological equipment shall be substantiated, or otherwise the updated metrological equipment justified in the design as well as the adjusted list and permissible values of the controlled parameters and the requirements for the instrumentation used in the course of testing shall be described.

In case of power density irregularity increase in comparison with the initial design the location of

additional measurement points for enhancement of the in-core measurement accuracy and the adjusted procedure for computational reconstruction of the power density field shall be substantiated.

In case of necessity administrative and technical arrangements for the ICMS refurbishment (including the software applications of the ICMS) shall be specified.

4.2.3. Tests and inspections

Testing programs and methods for the nuclear core and its assemblies, the non-destructive control and testing methods confirming the design characteristics of the nuclear core assemblies shall be described; the list of regulatory documents defining the requirements for the control and testing scope and methods shall be presented. The incoming control programs for the nuclear core assemblies at the NPP, the acceptance certificate of the Inter-Departmental Commission, the list of nuclear-hazardous works with the core and its assemblies shall be provided.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the methodologies and programs of the irradiation and post-irradiation testing of fuel assemblies with the new fuel type shall be presented.

The hardware and methods provided in the design for leak-tightness control of the fuel element claddings, particularly for the fuel elements made of the new fuel type, on the shut-down and (or) operating reactor aimed to ensure reliable and timely detection of leaking fuel elements shall be described. Methods used to monitor leak-tightness of the fuel element claddings in the shut-down and (or) operating reactor shall be provided and substantiated.

4.2.4. Design analysis

4.2.4.1. Normal operation

Functioning of the nuclear core and fuel assemblies under normal RP operation conditions, including power rising to the MCL and transient modes in case of scheduled start-ups and shutdowns shall be described. The state of the nuclear core in these modes, interaction with other RP systems during performance of the specified functions shall be demonstrated.

4.2.4.2. Safe operation limits and conditions

Safe operation limits for the nuclear core components shall be specified. The reference to the RP design documents and the NPP SAR sections containing substantiation of these limits shall be given.


- the limit for the fuel (according to temperature or absence of melting);

- the limits for the fuel element claddings (according to temperature and density);

- the limits for the nuclear core (according to reactivity and power change period). The thermal power limit for the nuclear core (the power value when the limit for the fuel element cladding temperature or the fuel temperature can be reached in the course of a design basis accident transient process).

Actuation of the emergency protection shall be provided upon reaching of the safe operation limits. The setpoint values shall be specified, and availability of the sufficient margin from the setpoint to the limit value shall be demonstrated.

The safe operation limits with regard to the nuclear core state shall be provided: according to the

specific loading of fuel elements, the coolant boiling, the coolant activity, the power-to-flow ratio and other limits established in the RP design.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the relevant safe operation limits and conditions, particularly for the fuel element damage, shall be specified. It is necessary to specify any potential additional arrangements provided in the design in order to maintain the ratio between the activity of fission products in the primary circuit coolant and the fuel element damage limit as established in the design.

4.2.4.3. Nuclear-hazardous works

The list of nuclear-hazardous works for handling of the nuclear core assemblies inside the reactor plant and in case of complete core unloading (if such an operation is provided in the design) shall be presented.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type applicability of the existing list of nuclear-hazardous works shall be confirmed, or the updated list shall be provided.

4.2.4.4. Design substantiation

Information on any works performed in order to substantiate the nuclear core and FA design shall be provided; this information shall be divided into the following groups:

- neutron and physical substantiation (provided in par. 4.2.7);

- substantiation of thermohydraulic characteristics (see par. 4.2.8);

- strength substantiation.

Information on the S&RW and R&DW performed to substantiate the nuclear core design shall be presented in accordance with the following scheme:

- the list of experimental works, S&RW and R&DW, including the ones performed at operating NPPs;

- description of the experimental methods;

- analysis of the experiment results.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the scope of additional stand-based and reactor experiments justified in the design and aimed to substantiate safety of the new core loadings with such fuel shall be presented.

4.2.4.5. Functioning in case of any failures

The list of initiating events and analysis of failures of the reactor and the primary circuit systems including human errors shall be presented, and their impact on the RP operability and safety shall be assessed.

Common cause failures shall be considered in the analysis of failures, and qualitative (in case of necessity) and quantitative assessment of their consequences shall be provided.

Impact of these failures on the operability of the reactor, the primary circuit system and other RP systems shall be analyzed. The list of systems and equipment required to mitigate and (or) to eliminate consequences of these failures shall be provided.

The section shall also include the list of all design basis accidents (the reference to Section 15 may be given) and the list of beyond design basis accidents considered in the design (also with the reference to Section 15).

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the reviewed list of design basis accidents and the list of beyond design basis accidents considered in the design shall be presented with due regard for peculiarities of the new fuel types that shall be analyzed in Chapter 15.

4.2.5. The reactor shutdown system - the CPS control rods

4.2.5.1. Purpose and functions of the system

Classification of the CPS control rods in accordance with their functional purpose (PSS), the safety class of the components and the seismic category, as well as the classifcation designation shall be specified.

Information on the regulatory basis of the reactor shutdown system design shall be provided.

4.2.5.2. Design basis

Information on the design basis (efficiency, response time) for normal operation conditions and accidents shall be provided.

4.2.5.3. Description of the CPS CR design

Description of the CPS CR design with indication of the purpose for the main components and information on the CPS CR groups shall be provided.

Design and purpose of the CPS CR guiding channel - the CPS sleeves shall be described, including the CPS CR drawings with the basic physical dimensions and position of the rods in relation to the nuclear core.

Operability of the CPS CR shall be confirmed by experience of operation in other reactors and stand-based tests.

The basic design characteristics of the rods shall be specified.

4.2.5.4. Materials

The information provided in par. 4.2.1.1 shall be used. Any sources of operability confirmation for the materials of the CPS CR and the CPS sleeves shall be indicated.

4.2.5.5. Quality assurance

Information on the NPP QAP for manufacturing of the rods shall be provided.

4.2.5.6. Tests and inspections

Frequency of control and the list of the controlled parameters for the CPS CR used to determine the criteria of operability loss (decrease of physical efficiency below the certain level, absence of any rod movements) shall be provided and substantiated.

The list of S&RW and R&DW performed in order to substantiate the CPS CR design and operability, particularly for manufacturing and physical weighing of the mock-up modesl, manufacturing and hydraulic testing of the mock-up models shall be provided.

4.2.5.7. Control and monitoring

The information provided in par. 4.2.2 shall be used.

4.2.5.8. Safe operation limits and conditions

The RP safe operation limits and conditions with regard to the CPS CR system state (response time characteristics, efficiency, permissible deviations from the vertical line, service life, periodicity of testing) shall be provided.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the relevant safe operation limits and conditions for the control and protection system shall be specified. Applicability of the existing setpoints for actuation of the preventive and emergency protections shall be also confirmed, or application of the new ones shall be substantiated.

4.2.5.9. Design analysis

4.2.5.9.1. Normal functioning

Functioning of the CPS CR under normal RP operation conditions and in case of any operational occurrences including design basis accidents shall be described, and information on the state of the CPS rods in these modes specifying the ways to determine and ensure their operability shall be provided.

4.2.5.9.2. Functioning in case of any failures

Any potential failures and damages of the CPS CR shall be analyzed with qualitative and (or) quantitative assessment of their consequences.

Information on any measures for prevention of failures or mitigation of their consequences adopted in the design of control devices and the CPS sleeves and in the course of their operation shall be presented. Failures of the equipment in the course of the CPS CR loading and unloading, in the refueling mode, failure to withdraw from the cell, unplanned rotation of the plugs shall be indicated.

Substantiation of the safe RP operation assurance in comparison with the operation results for the CPS CR of similar design and the results of stand-based tests and calculations shall be provided.

4.2.5.9.3. Design substantiation

Information on the works performed to substantiate the CPS CR design shall be provided:

- substantiation of the thermal and hydraulic characteristics;

- substantiation of operability (strength and reliability).

Information on each group of works shall include two parts - calculation and experimental. The calculation part shall in its turn include:

- the list of calculations;

- the applied methods and programs with information on their validation;

- results of the calculations and their analysis.

The experimental part shall consist of:

- the list of performed S&RW and R&DW;

- description of the applied methods;

- analysis of the experimental results.


- design efficiency value of the CPS CR with the relevant poison loading, efficiency decrease, burn-up, AE and CPS CR fluence within the specified service life;

- basic thermal and hydraulic characteristics of the CPS CR, particularly distribution of the coolant flow rate, temperature of the poison, AE claddings, parts of the rods and shroud tubes of the CPS sleeves, pressure difference in the rods and buoyancy force applied to them;

- basic strength characteristics of the CPS CR and the CPS sleeves determining their reliability, including the stress-strain state of the claddings and the CPS CR components, changes in the AE dimensions and shape due to swelling, creepage, temperature, interaction of the poison with the cladding, interaction of the AE bundle with the shroud tube, intreaction of the CPS CR parts with the shroud tubes of the CPS sleeves;

- the specified life limit, the specified service life and the specified shelf life of the CPS rods;

- criteria of the CPS CR operability loss.

In case of any refurbishment of the reactor nuclear core associated with usage of the new fuel type sufficiency of the existing reactor shutdown systems particularly the ones performing the functions of emergency protection with regard to their efficiency and fast response shall be confirmed, or the design materials for the refurbished systems shall be provided.

4.2.5.9.4. Design assessment

Compliance with the RD requirements shall be assessed.

4.2.6. Preventive emergency protection system


The data on presentation of the information related to the position of the PEP assemblies shall be given in the paragraph "Control and monitoring".

Compliance with the GSR requirements shall be demonstrated in the paragraph "Design assessment".

4.2.7. Neutron and physical calculation of the nuclear core

Information and analysis required to substantiate safe operation of the reactor nuclear core within its design service life under normal operation conditions and in case of any operational occurrences including accidents, as well as the information required to analyze the causes of accidents and to obtain the results presented in Section 15 shall be provided.

The information and analysis presented in this subsection shall be based on the design materials for the RP, the nuclear core, the nuclear core assemblies and on the S&RW results.

4.2.7.1. General description and basic neutron and physical characteristics of the nuclear core


- the NF type;

- peculiarities of the nuclear core design (configuration, techniques for the FA fixation, gaps between the fuel assemblies, side and end reflectors, characteristics of the structures behind the reflectors);

- the technique adopted in the design to level the power density field;

- the power control techniques adopted in the design;

- the CPS (EP) CR (see par. 4.2.2);

- presence of any other components in the nuclear core (experimental fuel assemblies, neutron source, etc.);

- the adopted techniques for reloading of fuel assemblies in the nuclear core, the CPS CR, PEP and blanket fuel assemblies;

- the list of the basic physical characteristics of the nuclear core and their values: the NF enrichment, the maximum power density, temperature margin before the NF melting under the rated conditions, efficiency of the CPS CR, the maximum reactivity margin, reactivity effects and coefficients, sub-criticality margins after fast shutdown of the RP, duration of the fuel campaign, the maximum fuel burn-up depth, the maximum neutron flux, the period between refuelings, residual heat curves for the nuclear core depending from the time after the reactor transfer to sub-critical state, etc.

4.2.7.2. Nuclear core operation modes in the course of the campaign


- general approach to organization of fuel replacement in the reactor;

- characteristics of the steady refueling mode;

- the list of the basic design nuclear core states in the steady mode;

- the main characteristics of the reloading programs for the fuel assemblies in the nuclear core, the lateral blanket and the CPS CR;

- general characteristics of the transient mode;

- general characteristics of the initial nuclear core (size, presence of diluent assemblies, etc.) and values of its basic physical parameters.

4.2.7.3. Characteristics of the power density field in the nuclear core and the adjacent structures

Information on distribution of the radiation field in the nuclear core and the adjacent structures in various states of the nuclear core characterizing the fuel campaign (before and after refueling, in the intermediate steady state and in any other states defined in the design), particularly the neutron fluxes in the nuclear core and the adjacent structures shall be provided.

4.2.7.4. Characteristics of the power density field with beyond-design positions of the CPS CR

The most unfavorable positions of the CPS CR shall be considered, and distribution of the power density fields and neutron fluxes for the selected configurations shall be provided.

4.2.7.5. Reactivity effects and coefficients related to changes of temperature and power

Temperature reactivity effects and coefficients adopted in the design and the structure of components of these effects shall be specified.

4.2.7.6. Reactivity effects related to changes in the nuclear core shape and dimensions

Investigation results for possible deformations of the nuclear core and its components occurring in the course of the RP operation with the rated power and in transient modes, as well as the values of reactivity effects occurring in case of such deformations shall be presented.

4.2.7.7. Reactivity effects related to sodium density changes

Values of sodium density in different temperature states of the nuclear core shall be provided: values of the reactivity effects related to sodium density changes in different sub-zones of the nuclear core and in the reactor in general with uniform and non-uniform temperature distribution.

4.2.7.8. Doppler effect

Values of the reactivity effects caused by changes in the resonant interaction of neutrons due to temperature variations (Doppler effect) shall be presented. Doppler effect values shall be provided for different states of the nuclear core within the campaign, as well as by components - for the basic materials of the nuclear core and for different isotope compositions of the fresh fuel.

4.2.7.9. Asymptotic values of the temperature and power reactivity effects for different states of the nuclear core

Values of the temperature reactivity coefficient and its components shall be provided for different states with regard to fuel burn-up: temperature of the nuclear core components at the rated power, and also the power reactivity effect and its components for different states with regard to fuel burn-up.

4.2.7.10. Sodium void reactivity effect and other hazardous effects

Values of the sodium void reactivity effect (SVRE) for different RP sub-zones, different states of the nuclear core with regard to fuel burn-up and various isotope compositions of the fresh fuel; SVRE values for different possible scenarios of boiling propagation and sodium supersedure across the RP section, as well as results of the SVRE investigations on experimental stands shall be provided. Description of other hazardous effects shall be similar to the SVRE description.

4.2.7.11. Efficiency of the control devices

Efficiency of the control devices shall be specified for different states of the nuclear core with regard to fuel burn-up. Efficiency of the control devices shall be considered for both individual devices and groups of devices and the entire control system with due regard for interference.

Efficiency values for the control devices depending on the poison burn-up shall be also specified.

4.2.7.12. Reactivity changes due to fuel burn-up Neptunium reactivity effect

The reactivity change value due to fuel burn-up, components of this effect for various isotopes and different reactor sub-zones and for different nuclear core states with regard to fuel burn-up as well as the neptunium reactivity effect value shall be provided.

4.2.7.13. Reactivity balance Compliance of the reactivity characteristics with the requirements of the NPP RF NSR

Analysis of the reactivity balance and compliance of the reactivity characteristics with the requirements of the NPP RF NSR shall be provided. The reactivity balance shall be developed with due regard for any potential errors in determination of reactivity effects. Errors shall be defined on the basis of the calculation analysis of the experimental values obtained on the model assemblies and operating reactors.

4.2.7.14. Analysis of the reactor sub-criticality in the course of refueling Neutron source, location and sensitivity of neutron detectors, control of sub-criticality


- general approach to control of the RP sub-criticality;

- the neutron source, its design and basic characteristics;

- neutron background of the nuclear core depending on the isotope composition of fuel and its burn-up degree;

- location and sensitivity of the neutron detectors;

- requirements for the refueling control and compliance with these requirements in the design under consideration.

4.2.7.15. Power monitoring

The applied neutron detectors and their characteristics for the RP power measurement shall be described in brief. Compliance of the selected power measurement system with the requirements of the NPP RF NSR and the capability of the power measurement system to control any distortions of the power density field caused by beyond-design positions of the control devices and any other reasons shall be analyzed.

4.2.7.16. Applied methods, programs and constants for physical calculations

Brief description of the programs and constants used for physical calculations shall be provided. The validated programs as well as degree of preparation for validation of any other applied programs shall be specified: availability of verification reports, user manuals and other documents.

In case of any refurbishment of the nuclear core related to usage of the new fuel type verification and validation results for the methodologies and codes used to determine neutron and physical characteristics of the nuclear core with the new fuel type shall be specified with due regard for analysis of uncertainties.

4.2.7.17. Basic results of experimental RP physics investigations on critical assemblies and operating fast reactors

Description of the model critical assemblies and the list of experiments performed on these assemblies shall be provided. The basic results of the calculation analysis of the experiments and transfer of these analysis results for assessment of error in physical characteristics of the RP design shall be presented.

In case of any refurbishment of the nuclear core related to usage of the new fuel type information on all neutron and physical characteristics of the nuclear core with the new fuel type specified in Section 4.2.7 shall be provided.

4.2.8. Thermohydraulic calculations

4.2.8.1. Design restrictions

Information on any design restrictions affecting the thermal and hydraulic characteristics, design modes of the RP and selection of its parameters shall be provided. They shall include:

- maximum temperature of the fuel element claddings;

- maximum temperature of the coolant;

- the coolant temperature change rate;

- maximum linear load on the fuel elements;

- maximum coolant flow velocity in the nuclear core and the margin to the FA emersion;

- the RCP net positive suction head;

- permissible limits for the sodium level change in the reactor.

4.2.8.2. Thermohydraulic calculation of the nuclear core


1. Distribution of the coolant flow and linear power density

The following shall be described:

- the scheme of the throttling zones in the nuclear core and the lateral breeder blanket (if any);

- distribution of the coolant flow rates by the throttling zones through the inter-cassette spaces and for the RP vessel cooling (with due regard for the potential different number of operating loops);

- average and maximum values of linear power density for different enrichment zones and throttling zones in the beginning and in the end of the campaign;

- the coolant temperature at the outlet of the nuclear core and the entire RP with due regard for the coolant flow rate distribution in the beginning and in the end of the campaign;

- temperatures of the fuel element claddings at the oultet of throttling zones with due regard for any potential non-uniformities of temperature distribution.

2. Pressure differential in the nuclear core and flow resistance

The scheme for the coolant flow arrangement at the reactor inlet (for example the high and low pressure header) shall be described, the pressure differential values in the nuclear core and on the lateral blanket and the corresponding flow resistance distribution by the components of the nuclear core and the lateral blanket shall be provided.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type and any design difference of the fuel assemblies with new fuel from the regular fuel assemblies their thermal and hydraulic compatibility shall be confirmed.

3. Methods and calculation programs

Information on any methods and calculation programs used in thermohydraulic calculations of the nuclear core, information on their verification or reliability substantiation for the obtained results with due regard for analysis of uncertainties shall be provided.

Information on any experimental works performed to substantiate the applied methods and calculation programs shall be presented.

Information on accuracy of the obtained thermohydraulic calculation results shall be provided.

4.2.8.3. Thermohydraulic calculations of the RP

The thermohydraulic calculations for the primary and secondary circuit of the RP and the emergency heat removal system shall be described.

The description shall include the following information.

1. Data on configuration of the equipment and pipelines of the RP circuits

The thermohydraulic diagram of the RP shall be provided:

- the number of the coolant circulation circuits and their purpose (the normal heat removal system, the emergency heat removal system);

- the type of the coolant movement activator (forced circulation, natural circulation);

- the list of equipment and pipelines in each circulation circuit, design values of the coolant flow rate for each circuit component and pressure differentials at the corresponding flow rates;

- the coolant circulation schemes for each circuit, elevations of the loop components (equipment, pipelines) for different circuits, their geometrical characteristics (particularly the coolant circulation path length in the component), the coolant volume in each component;

- the coolant levels in the RP circuit components and pressure of the gas medium in the design modes.

2. Design operation modes for the RP

The section shall contain the following:

- the list of design modes (with the reference for the corresponding subsection of Section 4);

- thermal and hydraulic peculiarities of each design mode;

- the coolant parameters and their change rates in various design modes;

- the coolant temperature distribution in different RP circuits in the design modes.

3. Methods and calculation programs

Information on any methods and calculation programs used in thermohydraulic calculations of the reactor plant, information on their verification or reliability substantiation for the obtained results shall be provided.

Information on accuracy of the obtained thermohydraulic calculation results shall be provided.


Programs and methods for tests and inspections to be used in order to confirm design thermal and hydraulic characteristics of the nuclear core and the RP circulation circuits shall be described.

4.2.9. The CPS actuators

The contents of this section shall be based on the developed design documentation for the CPS actuators, the RD requirements applied to the CPS actuators, the developed QAPs, operation experience of the prototypes, testing of the pre-production models and reports issued in the course of S&RW and R&DW and shall comply with the structure given below.

4.2.9.1. Purpose and design basis


- information on the composition, purpose and functions of the actuators;

- classification of the actuators in accordance with safety and seismic resistance;

- criteria, principles and design limits for the actuators under normal operation conditions, in case of any operational occurrences and design basis accidents;

- permissible limits for the main mechanical and strength characteristics and permissible values of reliability parameters for the actuators.

4.2.9.2. Description of the design


- description of the actuator designs with indication of the individual devices (components) performing independent functions including the control, fastening and sealing devices;

- sufficiently detailed drawings and schemes illustrating the design, kinematic operation diagrams and location of the actuators;

- basic technical characteristics of the actuators;

- the list of systems and equipment affecting the functioning of the actuators.

4.2.9.3. Materials

Information on the grades and properties of steels and materials used in the actuators as well as substantiation of their operability within the required time period in the liquid-metal sodium medium under the design temperatures and radiation exposures corresponding to the normal RP operation conditions and any operational occurrences including design basis accidents shall be specified.


References to the QAPs for development (design), manufacturing, acceptance and installation of actuators shall be given, and the main requirements stipulated in these programs and regulatory documents governing the requirements to quality assurance for the actuators and their assemblies.

4.2.9.5. Control, monitoring and testing


- principles for control of the actuators and monitoring of their state;

- characteristics of the control signals for the actuators;

- analysis of any potential controlling actions of automation devices and operators on the actuators;

- methods, means, scope and frequency of state control and testing for the actuators in order to ensure their operability in the course of operation and compliance with the regulatory requirements;

- information on the commissioning works for the actuators, including the list of testing programs demonstrating sufficiency of pre-operational tests for the actuators in order to substantiate safety of the RP operation and the list of arrangements aimed to prevent accidents in the course of testing.




- description of functioning of the actuators under normal RP operation conditions including transient modes during scheduled start-ups, power changes and shutdowns;

- description of the state of the actuators, their interaction during performance of the required functions;

- requirements for reliability and safety of the safety-related systems and equipment interacting with the actuators;

- description of functioning in case of any failures of the equipment actuators and systems and characteristics of the measures provided in the design to ensure functioning of the actuators in case of above-mentioned failures.



- analysis of failure consequences for the actuators, including failures due to human errors;

- description and sufficiency substantiation for the measures aimed to prevent common cause failures of the actuators including internal and external impacts and failures of systems and equipment;

- qualitative and quantitative (in case of necessity) assessment of failure consequences, particularly characteristics of the changes in the basic RP parameters affecting safety;

- the list of failures of the actuators representing initiating events for operational occurrences including design basis accidents that require additional analysis in the relevant section of the RP safety analysis report.


It should be demonstrated that the actuators comply with the safety RD, are tried out in operation of fast reactors or tested under the conditions closed to the required ones and substantiated by S&RW and R&DW.


Compliance of the actuator design with the RD requirements shall be assessed.

4.2.10. Reactor vessel



- information on the purpose and functions of the reactor pressure vessel;

- classification of the reactor pressure vessel in accordance with its safety impact and seismic resistance;

- regulatory basis of the design;

- criteria, principles and design limits used as the basis for the reactor pressure vessel design for

normal operation conditions and any operational occurrences including design basis accidents;

- the list of the reactor vessel failures considered in the NPP safety analysis.



- description of the reactor pressure vessel design with indication of the individual components performing independent functions including the control, fastening, sealing, warm-up and drainage devices;

- drawings and diagrams illustrating the design;

- the main technical characteristics of the reactor vessel;

- information on the adopted technical solutions aimed to prevent formation of sodium oxides and sodium compounds on the equipment surface and ingress of these deposits into the nuclear core.

4.2.10.3. Materials

The list of regulatory documents containing the requirements for the applied materials and information on the steel grades and properties for the reactor pressure vessel, substantiation of their capability to operate within the RP service life in the liquid-metal sodium and argon medium under the design temperatures, temperature variations and radiation exposures corresponding to the normal RP operation conditions and any operational occurrences including design basis accidents shall be provided.



- methods, means, scope and frequency of metal state control for the reactor pressure vessel in order to ensure its operability in the course of operation and compliance with the regulatory requirements;

- results of the stress-strain state determination for the pressure vessel material in the course of the RP start-up and commissioning.

4.2.10.5. Tests, inspections and metal state control

Information on the testing of the blanks for the reactor pressure vessel in the course of manufacturing, the incoming control of the reactor vessel or its constituent parts prior to installation, in the course of control during installation, testing for strength, leak-tightness and stability subsequent to installation shall be provided.




- description of the reactor vessel functioning under normal operation conditions in all modes prescribed in the operating regulations for any potential combination of loads (thermal, cyclic, seismic, impact, vibration, radiation, corrosion, etc.);

- analysis of any potential failures of the reactor vessel components with assessment of their consequences based on the PSA;

- compliance with the established requirements for the mechanical, strength and reliability characteristics of the reactor pressure vessel in all functioning modes.



- analysis of the consequences of any failures of the reactor pressure vessel or its components;

- the list of the reactor vessel failures representing initiating events for any operational occurrences including design basis and beyond design basis accidents and requiring additional analysis in the relevant section describing the RP safety analysis.


Compliance of the reactor pressure vessel with the regulatory requirements, application of the basic constructive solutions, experience in manufacturing, installation, testing and operation of the pressure vessels at similar operating plants as well as substantiation of the design with the documentation issued in the course of S&RW and R&DW shall be demonstrated.

4.2.10.6.4. Safe operation limits

The following limits shall be specified for the reactor pressure vessel:

- pressure;

- temperature;

- irradiation;

- strength.

4.2.10.6.5. Maintenance and maintainability

Information on maintenance and repair of the reactor pressure vessel and brief description of the repair processes shall be provided.

4.2.10.6.6. Reliability analysis for the RP vessel

Information on reliability analysis and the design failure probability values for the reactor vessel shall be presented.

Distribution of the neutron flux and fluence at the nuclear core boundaries and on the walls of the reactor pressure vessel for various periods of the reactor service life shall be provided.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the radiation resistance of the reactor vessel shall be additionally substantiated, and restrictions with regard to fast neutron fluence on the reactor vessel and the reactor internals shall be determined.

4.2.10.6.7. Control and monitoring


The list of measurement points and information on the diagnostic systems shall be provided.


Compliance of the reactor vessel design with the requirements and principles of safety and

feasibility of the design solutions shall be assessed.

4.2.11. In-vessel handling equipment

Information shall comply with the requirements for description of the RP systems.

4.2.12. Reactor cavity

Information on the composition of the cavity components and their characteristics shall be presented.

4.3. Primary circuit systems

The flow diagram of the primary circuit with indication of it main equipment and boundaries of the elevations shall be provided. In this case description of the primary circuit coolant circulation system, information on the coolant flow rate distribution within the pressure space, components of the total flow resistance of the primary circuit and distribution of the coolant temperatures among the primary circuit shall be specified.

Information on the components and systems included into the primary circuit shall be presented. It shall be sufficient to assess their impact on the NPP safety in general and shall include the purpose of the components and systems, design criteria, indication of the group, safety class and seismic category they refer to, characteristics and description of the design, assessment of compliance with the adopted design criteria.

The counterpart of the component (system) with well-known operation experience shall be specified.

4.3.1. Reactor coolant pump

Description of the RCP itself, the electric motor, the control, monitoring, protection and interlocking systems, the check valve with the drive, the design criteria and substantiation of their fulfillment shall be provided. Description of the RCP itself shall define the design operation modes for both power operation and shutdown of the RP; conditions for reliable functioning of the pump with various rotation speeds (in case it is adjustable) and different combinations of the operating pumps (absence of cavitation, motor overloading, etc.). The basic assemblies of the RCP shall be described within the scope sufficient to assess their impact on the pump operability and safety of the RP in general. Description of the RCP instrumentation shall be brief, but shall contain the list of protections and interlocks restricting the operation conditions as well as analysis of their impact on the RP safety.

Potential failures of the pump systems and components and their impact on the RP safety shall be considered in order to detect any failures requiring special analysis in Section 15.

In the course of the RP safety analysis in case of any failures or malfunctions of the RCP with due regard for peculiarities of the particular plant attention shall be paid to confirmation of compliance with the following requirements:

- maintenance of the required sodium flow rate through the nuclear core during the RP power operation and shutdown;

- elimination of simultaneous RCP failures (common cause failures);

- minimization of the possibility for the pump shaft blockage and any risk of damage for the

structures of the RP or the primary circuit;

- gas capture elimination;

- prevention of oil ingress from the pump system to sodium;

- assurance of the required pump run-out duration even in case of a fire (sodium or ordinary) on the RP roof;

- assurance of the guaranteed minimal flow rate for natural circulation with the pumps shut down;

- minimization of fire hazard caused by the pump itself or its systems;

- prevention of any argon leakages from the gas space of the primary circuit along the pump shaft or devices for the pump installation on the RP;

- prevention of the RCP rotation speed increase.

Information on control of the RCP rotation speed, description of the check valve and its control system shall be provided. Analysis and conclusions shall be confirmed with the relevant calculations, experiments or operation experience.

4.3.2. Intermediate "sodium-sodium" heat exchangers

Description of the intermediate heat exchangers shall be given in Section 5. Only the information on the IHE functions shall be provided in par. 4.3.2.

It should be demonstrated that the basic safety functions of intermediate heat exchangers are:

- heat removal from the primary circuit in all normal operation modes as well as in emergency modes in case residual heat removal is arranged via the SG or the air heat exchanger connected to the secondary circuit. It should be demonstrated that emergency heat removal will be ensured even in case of any leakage from the RP vessel;

- reliable separation of the radioactive primary circuit and non-radioactive secondary circuit.

It should be demonstrated that the structural components of air heat exchangers will not break in case of any pressure increase in the secondary circuit caused by the maximum water leakage to sodium in the SG, and the structural material of the AHE will remain stable in the sodium-water medium within the time required for the secondary circuit loop cleanup.

4.3.3. Protective gas pressurizing system

Information shall comply with the requirements for standard content of the protective gas pressurizing system.

4.3.4. Primary circuit pipelines (pressure header)


Purpose and design basis for the RP pressure header and its basic characteristics shall be described.


- information on the purpose and functions of the RP pressure header;

- classification of the RP pressure header in accordance with its safety impact and seismic resistance;

- criteria, principles and requirements used as the basis for the RP pressure header design;

- the list of the RP pressure header failures that shall be taken into account in the NPP safety analysis.



- description of the RP pressure header design with indication of individual components performing independent functions;

- drawings and diagrams illustrating the design;

- basic technical characteristics of the RP pressure header.

4.3.4.3. Materials

The list of regulatory documents governing the requirements for the applied materials and information on the steel grades and properties for the RP pressure header, substantiation of their capability to operate within the RP service life in the liquid-metal sodium medium under the temperatures and radiation exposures corresponding to the normal RP operation conditions and any operational occurrences including design basis accidents shall be provided.


Description of the NPP QAPs for development, manufacturing, acceptance, installation and testing of the RP pressure header, the main requirements specified in these programs and the regulatory documents governing the requirements for quality assurance of the RP pressure header shall be provided.

4.3.4.5. Tests and inspections including in-service ones

Information on testing of the blanks for the RP pressure header in the course of manufacturing, control in the course of installation and strength testing shall be specified.






- description of the RP pressure header functioning under normal operation conditions in all modes under any potential combination of loads (thermal, cyclic, seismic, impact, hydraulic, vibration, radiation, corrosion, etc.);

- compliance with the established requirements for strength, deformation and leak-tightness of the pressure header in all functioning modes.



- analysis of the consequences of any RP pressure header failures;

- qualitative and quantitative (in case of necessity) assessment of the failure consequences;

- the list of the RP pressure header failures representing initiating events for any operational occurrences including design basis accidents and requiring additional analysis in the relevant section describing the RP safety analysis.


It should be demonstrated that the RP pressure header complies with the regulatory requirements, is substantiated with the operation experience for the pressure headers at similar operating reactor plants and studied in the course of S&RW and R&DW.


Compliance of the RP pressure header design with the regulatory requirements shall be assessed.

4.3.5. Primary circuit sodium purification system


Purpose of the system and requirements for the sodium coolant purity maintenance shall be described.

The list of limited impurities and restrictions on their content in sodium shall be specified. Depending on the impurity type it is necessary to specify the means for maintenance of this impurity content below the established limit as well as any instruments and devices enabling to determine the content of the limited impurities with sufficient accuracy and sensitivity.

The design principles shall be stated; references to the documents, standards and rules to be taken into account for development of the system shall be given.

The class of the system components in accordance with their safety impact shall be specified. All pipelines and equipment of the system shall refer to safety groups in accordance with the RD requirements. Seismic category of the system shall be determined.

4.3.5.2. Description of the system

The purification system and its main units and equipment shall be described. Components performing independent functions shall be indicated. Performance of the functions under normal operation of the system and in case of any malfunctions shall be demonstrated. The possibility of access to the equipment for the purpose of repair and maintenance shall be demonstrated.

The system configuration (units, equipment) with indication of the number and basic process characteristics shall be presented.

The systems supporting normal operation of the purification system (the cooling circuit, heating, etc.) shall be described in brief, and interfaces of the purification system with any other systems (the primary circuit, the tank farm, etc.) shall be specified.

The flow diagram of the system with indication of its boundaries, all components and

measurement points for the main parameters shall be provided.

4.3.5.3. Materials

Information confirming compliance of the structural materials, manufacturing and control methods with the RD requirements, and also information on the coolant compatibility with the system materials shall be provided.


Information on the QAPs for manufacturing and installation of the equipment and systems shall be presented.

4.3.5.5. Tests and inspections Commissioning works

Information on the commissioning program for the purification system (adjustment and commissioning of the system equipment, the electric heating systems for the equipment and pipelines, the valve control systems, the process parameter monitoring systems) shall be provided.

4.3.5.6. Control and state monitoring for the system in the course of operation

Process control shall be provided. The instrumentation scheme, the list of measuring equipment for the system parameters shall be presented. In-process (plug indicators for impurities, electrochemical cells) and out-of-process (sodium sampling with subsequent chemical analysis) monitoring means shall be provided in order to determine the coolant quality.


4.3.5.7.1. Functioning of the system under normal operation conditions

Functioning of the system under normal operation conditions shall be described.

4.3.5.7.2. Functioning of the system in case of failures

Functioning of the system in case of failures shall be described. A beyond design basis accident with complete cross-section rupture of a sodium pipeline without safety shell is considered in Section 15. The maximum sodium spillage without any interference of the operating personnel should be assessed.

4.3.5.8. Reliability analysis

Reliability analysis for the system shall be performed with due regard for the experience in operation of similar systems at operating NPPs and based on the quantitative analysis.

5. REQUIREMENTS FOR THE SECTION "SECONDARY CIRCUIT AND THE ASSOCIATED SYSTEMS"

Information on the systems and components of the secondary circuit shall be provided including:

1. Main equipment, pipelines, systems:

- intermediate heat exchangers of the primary and secondary circuits;

- the secondary circuit RCP;

- SG;

- the secondary circuit pipelines;

- the protective gas pressurizing system;

- the coolant purification system (within the secondary circuit boundaries);

- the secondary circuit RCP cooling system;

- the control system;

- the EPS of the SG and the secondary circuit (described in Section 12);

- valves of the secondary circuit: gate valves, fast-response shut-off valves, PSE for sodium, fast-response relief valves of the SG EPS.

5.1. General characteristics of the design

5.1.1. Purpose and design basis

The purpose of the secondary circuit, information on the design of the secondary circuit and its main systems shall be provided. Classification of the secondary circuit equipment in accordance with the CSR, categories stipulated in the rules and regulations for seismic-resistant nuclear power plants and quality groups in compliance with the NPU Rules shall be provided.

The section shall include the list of regulatory documents on safety containing the requirements the secondary circuit shall comply with and the basic design principles.

5.1.2. Description of the process flow diagram

Description of the secondary circuit system with indication of the components and systems performing independent functions shall be provided, and the safety functions performed by these components and systems shall be specified.

References to other sections of the NPP SAR containing more detailed requirements for individual systems and components of the secondary circuit shall be given. Compliance with the requirements of the NPU Rules and other regulatory documents shall be confirmed.

Brief description of any support and suspension components installed on the pipelines and equipment in order to accommodate seismic loads shall be provided.

The possibility for the coolant drainage and absence (presence) of any dead zones as well as the possibility to fill the system with the coolant and to remove gas from it (through the vapor traps) shall be demonstrated.

The flow diagram of the secondary circuit with indication of the secondary circuit boundaries and all main components shall be provided in the flow diagram description.

All systems connected to the secondary circuit and the method of their disconnection from the secondary circuit as well as elevations of the equipment and pipelines included into the secondary circuit shall be indicated on the diagram.

The diagram shall have explanations with indication of the the coolant temperature and volume for each section, flow passages of the pipelines, the operating pressure and flow rates in the circuit in steady plant operation modes.

Routing of the secondary circuit within the building in isometric image shall be provided, or the reference to a separate document containing the above-mentioned routing shall be given. The diagram shall have explanations on the procedure for filling and drainage of the coolant as well as indication of the coolant drainage and pressure relief routes in case of any emergencies associated with loss of leak-tightness in the secondary circuit.

The scheme for monitoring of the parameters for the secondary circuit and the connected non-isolable systems located within the secondary circuit boundaries shall be provided in description of the flow diagram.

Descriptions of the flow diagrams for the associated systems of the secondary circuit (the ERCS, the SG EPS, the purification system, etc.) shall be considered in the relevant sections of the SAR.

The drawings shall have explanations and requirements for the NPP rooms related to separation of the loops, indication of the room categories, any potential coolant spillage volumes in case of leak-tightness loss in the pipelines and equipment of the secondary circuit.

Detailed description of the design solutions for detection and confinement of any sodium leaks from the secondary circuit systems (detection of the leak point, isolation of the leaky section, activation of the fire extingushing systems) shall be provided. It should be demonstrated how the probability of the pipeline rupture is minimized; in this case the following shall be presented:

- design dimensions of the pipeline defects and dynamics of their development, the most probable rupture points;

- analysis of the pipeline breakage consequences (the coolant spillage volume, impact of the coolant loss on the circuit operation, consequences of the burning sodium exposure);

- protection against pipeline breakage consequences (confinement and limitation of the coolant spillage, sodiem combustion suppression).

The following shall be demonstrated:

- the secondary circuit is designed in such a way so that to ensure access to the equipment for inspections, maintenance and repair;

- the exposure doses for the workers do not exceed the established regulatory limits.

It should be demonstrated that all systems and components of the secondary circuit are designed with due regard for the capability to withstand the environmental conditions (pressure, temperature, humidity, radiation) occurring in the course of normal operation, in case of any operational occurrences, pre-emergency situations and design basis accidents within the entire service life, and any failures of the systems and components not referred to the first seismic category will not result in failures of the systems and components referred to the first seismic category.

The design solutions shall be confirmed by the existing operation experience for the equipment and pipelines.

5.1.3. Quality assurance

The list of effective QAPs for the relevant SAR development stage shall be presented.

Information on the application scope for particular QAPs shall be provided.


The list of all measurement points and controlled elements shall be presented. Control of the secondary circuit as well as characteristics of the parameters (setpoints) for activation of the process protections and interlocks; requirements for the accuracy of the measured parameters, metrological support for the measuring means and techniques shall be described.

The design requirements for monitoring of the equipment and pipeline integrity (based on the requirements of the NPU Rules and other regulatory documents) shall be specified. In case the design requirements are established by the designer their establishment shall be substantiated. The control methods adopted by the designer in addition to the ones prescribed in the regulatory documents shall be described. The methods for registration of initial inspection results for the secondary circuit equipment and pipelines shall be described.

5.1.5. Tests and inspections in the course of operation

Completeness of in-service inspection and the required testing scope for the secondary circuit components shall be substantiated.

Substantiation shall contain:

- boundaries of the systems subject to control;

- location of the systems and components taking into account access for control purposes;

- control techniques and methods ensuring compliance with the RD requirements;

- control frequency;

- frequency and procedure for strength and leak-tightness testing. Compliance with the requirements of the NPU Rules shall be demonstrated.


5.1.6.1. Normal functioning of the secondary circuit system

Functioning of the secondary circuit system under normal operation conditions including transient modes in the course of scheduled start-ups and shutdowns shall be described.

State of the secondary circuit system and its components as well as their interaction during performance of the prescribed functions shall be described.

The main characteristics of the secondary circuit operation model shall be specified, including:

- name of the mode;

- the reactor power;

- design number of cycles within the service life and their characteristics (for example, the stress intensity range);

- duration of the regime;

- type of the regime;

- basic parameters of the regime for the secondary circuit (average temperature at the IHE inlet or outlet. protective gas pressure, the secondary circuit RCP rotation rate).

Reference to the documents containing substantiation of the operation model and the operation mode for the RP where the secondary circuit and associated systems are operated shall be given. Diagrams of time dependence of the secondary circuit parameters shall be provided for the main RP operation modes.

5.1.6.2. Functioning in case of any failures

The list of postulated initiating events and analysis of the secondary circuit system component failures particularly probability of their occurrence, including human errors shall be provided. Impact of their consequences on the system operability and safety shall be assessed.

Dimensions of design defects resulting in the coolant leakage in case of any design basis or beyond design basis accidents shall be provided and substantiated. Failures of the passive components (pipelines, intermediate heat exchangers, SGs, tanks, protective burst disks, etc.), active components (pumps, gate valves, valves, etc.), measuring means for the secondary circuit parameters shall be considered with analysis of common cause failures including potential fires.

Qualitative and quantitative (in case of necessity) assessment of consequences shall be provided for the considered failures.

Impact of these failures on operability of the secondary circuit and other RP systems shall be demonstrated. The list of systems and components required to eliminate consequences of these failures shall be provided.

Safe operation limits and conditions shall be specified.

5.1.6.3. Design substantiation

Information on the calculations performed in order to substantiate the design shall be provided.

It shall include:

- the list of all calculations performed;

- the list of methodologies and programs used to substantiate the design with indication of the application scope, assumptions, information on validation of the programs;

- calculation results.

Information on consideration of any counterpart operation experience shall be provided.

Information on the S&RW and R&DW performed to substantiate the design shall be presented in accordance with the following scheme:

- the list of all experimental works performed;


- results of the experiments with conclusions.

5.1.6.4. Design assessment

The data confirming compliance of the materials, manufacturing and control methods for the secondary circuit components with the requirements of the NPU Rules shall be provided.

The adopted manufacturing methods and compliance with the requirements of the NPU Rules shall be specified.

Compliance with the design basis stated in the section "Purpose and design basis" as well as with the RD requirements shall be demonstrated.

5.2. Secondary circuit systems and components

Information containing peculiarities of the individual secondary circuit systems and components shall be provided.

The component (system) counterpart with well-known operation experience shall be specified; any variations from the counterpart and the reasons to introduce them shall be indicated.

In case any component (or system) is completely taken from any other plants, or commercial products are used it shall be demonstrated that their technical characteristics, operation modes and conditions comply with the requirements for the plant under consideration.

The subsection "Materials" containing the list of technical specifications for metals used to manufacture the main components of the secondary circuit as well as welding and surfacing materials shall be provided.

In case the selected material is not specified or specified but used with any deviations the reference to the documents substantiating the possibility to apply the selected material shall be given.

Information on the applied welding types with the list of regulatory documents governing the requirements for welding shall be provided.

5.2.1. Intermediate heat exchanger of the primary and secondary circuit

The information shall be provided within the scope and in accordance with the procedure specified in Subsection 5.1. Besides the IHE characteristics presented in the section "Purpose and design basis" shall include the design coolant radioactivity limits in the secondary circuit of the IHE under normal operation conditions and substantiation of these limits.

The section "Materials" shall include the information on selection of the materials with due regard for peculiarities of the IHE and its manufacturing method affecting the requirements for the materials (for example, design and embedding technique for the heat exchanging tubes, etc.); it should be demonstrated how these peculiarities are taken into account in selection of the material (for example, selection of the material which does not require any heat treatment of weld seams due to the IHE operation conditions).

Compatibility of the IHE materials with the primary and secondary circuit coolant shall be also substantiated. The manufacturing technology for the main IHE assemblies shall be described in brief with the special focus on the manufacturing technology for the tube sheets, welding of complex weld joints; the embedding technique for the heat exchanging tubes shall be described. Techniques for the heat exchange surface cleaning in the course of manufacturing and cleanness control methods shall be described. Material selection for the heat exchanging tubes shall be substantiated, and requirements for the surface condition, heat treatment and other parameters important to ensure operability of the tubes shall be specified.

Description of the IHE transportation methods, measures provided in the design in order to prevent damage of any IHE components in the course of transportation and installation, the heat exchange surface preservation necessity and technique, preservation and cleanness control for the inner surface in the course of storage, installation and final assembly at the power unit shall be presented. The IHE installation procedure shall be described in brief.

The section "Control and monitoring" shall additionally include the arrangements provided in the IHE design in order to monitor the state of the IHE components in the course of operation in accordance with the requirements of the NPU Rules.

Consequences of any ruptures of heat exchanging tubes and other design basis accidents associated with the secondary-to-primary circuit leakages shall be also considered in the section "Functioning in case of failures", or references to the relevant sections of the NPP SAR where these situations are analyzed shall be given.

The ways to ensure compliance with the design criteria for prevention of unacceptable damage of the IHE heat exchanging tubes (due to vibration, corrosion damage, etc.) shall be demonstrated in the section "Design substantiation" and substantiated in the design. The following information shall be specified in the design substantiation:

- design operation conditions in the modes determinative for strength assessment of the heat exchanging tubes and point of their embedding to the tube sheets;

- results of calculations and experiments confirming that the adopted stress intensity level ensures reliable operation of the intermediate heat exchanger. References to the relevant sections of the NPP SAR may be given.

5.2.2. Main circulation pumps

The information shall be provided within the scope and in accordance with the procedure specified in par. 5.2.1.

The pump type and configuration shall be specified in the section "Purpose and design basis".

The necessity to connect the RCP to the reliable power supply system shall be defined.

The section "Description" shall include the information on the RCP components or references to the relevant sections of the NPP SAR. The basic technical characteristics of the pump shall be provided in the table with indication of the nominal rotation rate, power and the rotation rate adjustment ranges. The presented information shall include the brief description of the RCP auxiliary systems and their characteristics.

The section "Materials" shall include the information on selection of the materials with due regard for the RCP design peculiarities.

The lists and description of the instrumentation for the RCP and its auxiliary systems shall be presented in the section "Control and monitoring".

The section "Functioning in case of any failures" shall include the analysis results for any failures of the main pump components that can cause:

- loss of the secondary circuit integrity;

- loss of the coolant circulation in the secondary circuit loop;

- ingress of the RCP cooling medium into the secondary circuit coolant.

The coolant flow rate curves for the secondary circuit loop in case of the maximum permissible RCP power supply interruption and the RCP electric drive switch-off shall be provided.

Information on the RCP electric drive shall be presented in a separate sub-section. It shall include general descriptions and characteristics of the electric drive, the diagram and list of the protections and

interlocks, the main parameters of the supply network, the permissible power supply interruptions, potential failures (particularly the ones related to the rotation rate increase).

The information on the electric drive shall be provided within the scope and in accordance with the procedure specified in par. 5.2.1.

The rotation speed control system (in case any is provided in the design) for the secondary circuit RCP may be presented in Section 7.

The section "Auxiliary systems of the power unit" shall include the description of the auxiliary systems for the secondary circuit RCP:

- the RCP oil facilities including the oil supply system for the bearings of the pump and the electric motor, the oil supply system for the gas sealing of the pump shaft;

- the RCP cooling systems.

5.2.3. Steam generators


The assigned design basis for the SG design shall be provided in the section "Purpose and design basis".

The paragraph "Description of the design and flow diagram" shall include the SG configuration (the number of sections, modules and their purpose), configuration and purpose of the auxiliary systems (drainage, filling, gas blow-off, SG EPS, electric heating and heat insulation systems, the SG water circuit chemical flushing system, the SG piping system, the firing system, support structures, the SG diagnostic systems).

The general layout of the SG in the building and the SG process flow diagram shall be presented.

Peculiarities of the SG layout in the building eliminating any impact of an accident in one SG on other loops shall be described.

Description of the SG modules shall contain consolidated technical characteristics including the total heat exchange surface, the water and coolant volume in the module.

The technical data on the SG including the information on the third circuit shall be presented in a summary table.

The paragraph shall contain the following:

- analysis in comparison with the operated SG counterparts regarding thermohydraulic and structural parameters, schematic and constructive solutions, operation conditions determining the SG reliability;

- description of the SG transportation methods, measures provided in the design in order to prevent damage of any SG components in the course of transportation and installation, the heat exchange surface preservation necessity and technique, preservation and cleanness control for the inner surface in the course of storage, installation and final assembly at the NPP power unit as well as description of the SG installation procedure.

The section "Materials" shall include the information on selection of the materials with due regard for peculiarities of the SG and its manufacturing technique affecting the requirements for the materials

(for example, presence of the steam and water media interface, temperature pulsation, design and embedding technique for the heat exchanging tubes, etc.); it should be demonstrated how these peculiarities are used in selection of the material (for example, necessity to improve characteristics of the material in relation to its crack and corrosion resistance).

The list of structural materials used in the SG both for the secondary circuit and the SG piping (superheated and slightly superheated steam, feedwater) shall be provided.

Compatibility of the SG materials with the secondary and third circuit coolant shall be substantiated. The manufacturing technology for the main SG assemblies shall be described in brief with the special focus on the manufacturing technology for the headers, complex weld joints; the embedding technique for the heat exchanging tubes shall be described, and information on flaring of the heat exchanging tubes shall be presented. Techniques for the heat exchange surface cleaning in the course of manufacturing and cleanness control methods shall be described. Material selection for the heat exchanging tubes shall be substantiated, and requirements for the surface condition, heat treatment and other parameters important to ensure operability of the tubes shall be specified.

References to the QAPs shall be given in the subsection "Quality assurance".

The subsection "Commissioning tests" shall include the CW program and the list of systems supporting the SG start-up and adjustment.

The subsection "Control, monitoring and testing in the course of operation" shall include the list of the measurement points, process protections and interlocks, controlled elements, controllers and diagnostic systems. The arrangements provided in the SG design in order to ensure state monitoring for the components in the course of operation in accordance with the requirements of the NPU Rules shall be described.

Description of the SG operation model, the complete list of modes and the number of the SG working cycles, the procedure for maintenance, testing and repair of the SG as well as the safe operation conditions shall be provided in the subsection "Normal functioning of the system".

Consequences of any ruptures of SG heat exchanging tubes and other design basis accidents associated with the third-to-secondary circuit leakages shall be considered in the section "Functioning in case of failures", or references to the relevant sections of the NPP SAR where these events are analyzed shall be given.

Compliance with the design requirements for prevention of any unacceptable damages of the SG heat exchanging tubes (due to vibration, corrosion, etc.) shall be demonstrated in the subsection "Substantiation".

The following information shall be provided in the substantiation:

- design conditions in the modes determinative for strength assessment of the heat exchanging tubes and point of their embedding to the headers;

- results of calculations and experiments confirming that the adopted stress intensity level ensures reliable operation of the SG. References to the relevant sections of the NPP SAR may be given.

- evidence of compliance with the requirements of the NPU Rules for availability of the wall metal temperature monitoring and the coolant level indicators.

Information on the performed tests for the SG components and their brief results, the list of performed calculations confirming the SG operability and reliability shall be provided.

5.2.4. Pressurizing system


5.2.5. Emergency protection system for the steam generator

The SG EPS shall be presented in Section 12. Its description shall be provided in the same section. The description may be presented in Section 5.

The SG EPS is intended to ensure safe and reliable operation of the SG, to confine the products of sodium and water interaction in case of any emergency situations, water leakage to the secondary circuit coolant due to loss of leak-tightness in the SG heat exchanging tubes or tube embedding assemblies of the tube sheets.

The SG EPS usually includes:

- the indication system for water (steam) leakage to the secondary circuit;

- the system for fast disabling of the leaky (in the steam-water circuit) SG;

- the system for emergency discharge and confinement of the water-coolant reaction products;

- the SG EPS state monitoring system;

- the system for filling of the SG steam-water spaces with gas.


The following shall be additionally provided:

- methods and instruments for indication of leaks, the SG EPS scheme and the system configuration;

- description of the procedure for detection and confirnement of leaks and functioning of the system in general.

The following shall be specified for the postulated quantity of leaky SG tubes:

- leakage intensity in case of the maximum design basis accident;

- time for the coolant discharge from the SG section;

- steam pressure relief time;

- quantity of water (steam) entering the secondary circuit;

- sensitivity of the indication system;

- time for the alarm signal generation;

- the PSE actuating pressure value;

- the opening (closing) time for the fast-response valves.

The system peculiarities (for example, the PSE integrity check) shall be specified in the description

of commissioning, control and in-service inspection. Compliance with the requirements of the NPU Rules shall be confirmed.

Service life of the measuring and monitoring means applied in the SG EPS shall be specified.

The generalized algorithm for actuation of the SG EPS shall be presented.

The hydraulic shock value shall be assessed, or the reference to the hydraulic shock calculation in case of the SG shutdown and the SG EPS activation shall be given. In this case the design pressure in the individual secondary circuit components (the SG, the buffer tank, the secondary circuit pipelines, the IHE tube sheet) as well as the design pressure in the feedwater pipelines of the third circuit shall be specified.

5.2.6. Pipelines

Information on the set of pipelines subjected to the secondary circuit pressure in the course of operation shall be provided within the scope and in accordance with the procedure specified in par. 5.1.2. Information on the pipeline suspensions and supports as well as information on the auxiliary systems (drains, air vents, RCP-II overflow pipelines), the process components and systems (heat insulation and electric heating of the pipelines) shall be additionally provided.

The paragraph shall contain substantiation of the selected design scale for loss of the pipeline leak-tightness in case of any design basis and beyond design basis accidents and assessment of consequences of such leak-tightness loss, as well as references to the relevant sections of the NPP SAR.

5.2.7. Sodium purification system

Purpose and objectives of the secondary circuit sodium purification system are the same as for the similar primary circuit system.

Reference to the description of the primary circuit sodium purification system (par. 4.3.4) shall be given in the description, and any additions reflecting the peculiarities of the secondary circuit sodium purification system shall be provided.

5.2.8. Secondary circuit valves

Information on the shut-off, isolating and control valves shall be provided in accordance with the scheme given in par. 5.2.1. The information shall also confirm compliance with the requirements of the NPU Rules.

Information on the secondary circuit valves with indication of their characteristics (the flow passage, response time, etc.) and operating conditions (temperature, pressure, position, etc.) shall be presented.

5.3. System for emergency cooldown via the AHE

Description of the system shall be provided in Section 12. Mutual functioning of the ERCS via the AHE and the system of the secondary circuit pipelines shall be demonstrated in Subsection 5.3.

6. REQUIREMENTS FOR THE SECTION "THIRD CIRCUIT"

Information on the following process systems shall be provided:

- the turbine plant;

- the main steam pipeline system;

- the feeding and deaeration plant;

- the auxiliary steam system;

- the firing system;

- the main condensate system;

- the condensate purification system;

- the condenser vacuuming system;

- the condenser water pipeline system;

- the heating plant.

As the possibility of the RSb ingress from the RP to the turbine plant system is eliminated due to presence of the intermediate sodium circuit this part of the plant shall not be subject to analysis in the SAR to the same extent and in the same detail as other systems with important functions for safety assurance. The third circuit systems shall be classified according to their safety impact in compliance with the RD requirements. The sufficient scope of information enabling to get a general sense of the third circuit process systems shall be provided with the focus on the design and operation aspects affecting or capable to affect the RP safety.

This section shall contain the information on the calculation results confirming strength, stability and operability of the components in accordance with the classification for the components of each system. The necessity to take into account any natural or human-induced impacts in the strength calculations for the components of each turbine plant system shall be stated in the paragraph "Design basis".

6.1. Turbine plant

6.1.1. Purpose

Purpose of the turbine plant and its impact on the RP shall be described in brief.

The list of safety RDs containing the requirements the turbine plant shall comply with shall be provided. Initial data on the modes shall be specified.

Information on the equipment classification in accordance with the requirements of the regulatory documents on safety shall be provided.

The basic principles of layout solutions shall be stated.

6.1.2. Design basis

Information on the type of the turbine plant used in the power unit design shall be provided. Requirements for the turbine set cyclic load capability with indication of the permissible number of start-ups within the service life (cold start-up, hot start-up, scheduled and unscheduled shutdowns, load reduction to idle run, load reduction to the lowest limit of the adjustment range with subsequent loading; design duration of start-ups in various thermal conditions from steam supply to the turbine set

up to the rated load; the adjustment range of automatic power variation; deviation of the rotor rotation speed within the adjustment range and under emergency conditions) shall be specified.

The limit parameters characterizing impermissible exceedance of the turbine speed (for example, the maximum permissible rotor rotation rate in emergency modes, the time for reaching of the maximum permissible rotation rate) shall be specified.

Requirements for mitigation of any consequences caused by impact of any missiles due to mechanical damage of the rotor or blades as well as any short circuits in the generator on the NPP shall be provided.

The requirements for configuration of the turbine plant shall be specified.

6.1.3. Description of the process flow diagram

The flow diagram of the turbine plant, brief information on its design and the list of auxiliary systems shall be provided.

Information with substantiation of the turbine set layout and orientation, the locations of explosive and flammable materials shall be provided.

Information on the turbine plant as the source of missiles in case of any potential turbine overspeed or short circuits in the generator shall be presented. Information on any missiles that can be formed due to mechanical destruction of the turbine rotor or blades (the target areas, kinetic energy) shall be provided. It shall be demonstrated that any possible damages caused by missiles will not result in any disturbances of the SS functions, damage of oil systems, systems containing flammable gases or high-pressure gases.

The zones of potential missile ejection within the sector of +25 degrees in relation to the low-pressure cylinder rims of each turbine in the turbine hall shall be indicated on the layout plan of the turbine plant.

Description of the turbine plant operation modes (normal operation; functioning in case of any operational occurrences; functioning in emergency situations and in case of accidents) reflecting operation of BRU-A, BRU-K and other associated systems shall be presented. Information on functioning of the turbine and systems in case of any abnormal operation of the turbine plant itself or any malfunctions in the associated systems shall be provided.

Information on emergency modes of the turbine plant shall be specified.

Initiating events at the plant capable of causing any accidents shall be described. Complete analysis of accidents caused by failures of the turbine plant is presented in Section 15; reference to this section shall be given.

Brief description of the main turbine plant components and their classification shall be provided. Strength characteristics of the turbine disks and other most stressed devices shall be particularly presented (description and design of the power generator shall be provided in Section 8).

6.1.4. Materials used

Information on the materials used to manufacture the turbine parts (rotors, disks, working blades, vessels (heaters)) and data on the manufacturing technology shall be provided.

Information on the breaking strength of the high-pressure turbine rotor material shall be specified. The above-mentioned information shall be presented in the form of brief data at the NPP

PSAR development stage. More detailed information on the strength characteristics and the applied materials shall be provided at the NPP FSAR development stages.

6.1.5. Protection against impermissible overpressure

Information on substantiation of the selected means to protect the turbine plant against impermissible overpressure shall be specified.

6.1.6. Protection against overspeed

The system for protection of the turbine plant against overpressure as well as the redundancy methods, reliability assessment for the assemblies, the procedure for control and testing shall be described.

It should be demonstrated that the maximum possible rotor rotation speed will not result in any impermissible RCP rotation speed.


All measurement points required for generation of protections and interlocks, control and monitoring of the operation and emergency alarms shall be specified.

Protections and interlocks affecting the preventive protection of the reactor and the reactor EP or accelerated load relief shall be described.

Description of the turbine plant control systems shall be provided.

6.1.8. Tests and inspections

The program of pre-operational, commissioning and in-service testing of the turbine plant, its locking and control devices and the turbine plant safety trip shall be described in brief.


6.1.9.1. Normal operation

Normal operation modes for the turbine plant as well as operational occurrences causing transient processes in the RP shall be described in brief, particularly:

- sudden turbine plant load shedding;

- shutdown of the high-pressure heater;

- disconnection of the generator from the grid;

- disconnection of the generator with loss of vacuum in the condenser.

Impact of sudden load shedding and potential transient processes on feedwater supply to the SG and pressure maintenance in the SG shall be demonstrated. The operation of the turbine plant control system and its overspeed protection shall be presented. Operation of BRU-A, BRU-K and other associated systems particularly the superheated steam system, the auxiliary steam pipeline system, the firing system shall be reflected.

6.1.9.2. Functioning in case of operational occurrences

Information on the qualitative analysis of any potential failures of the turbine plant and its systems shall be provided.

Recovery of the turbine plant operation due to redundancy of the systems shall be demonstrated.

6.1.9.3. Functioning of the system in case of emergency situations and accidents

The list of initiating events in the turbine plant causing its accidents shall be provided.

Functioning of the system in case of these initiating events with due regard for operation of its components shall be demonstrated. It shall be demonstrated that any initiating event in the turbine plant will not result in an accident or emergency situation at the NPP.

The way to ensure cooldown and residual heat removal from the RP in case of these initiating events shall be demonstrated. Complete analysis of accidents at the NPP caused by an accident at the turbine plant is provided in Section 15.

6.1.9.4. Functioning under external impacts

The state (operation or shutdown) of the turbine plant in case of any external impacts (earthquake, blast wave, aircraft crash, tornado, etc.) shall be specified.

The level of external impacts when the turbine plant and the entire NPP should be stopped shall be indicated.

6.1.10. Reliability parameters

Reliability parameters for the turbine plant and its equipment shall be specified. At the NPP PSAR stage references to the basic design used as the basis for the reliability parameters shall be given; data on the reliability parameters may be provided based on the technical assignment for the turbine plant. Complete information on the above-mentioned materials shall be provided at the stage of NPP SAR.

6.1.11. Design assessment

The turbine plant design shall be assessed in accordance with its safety impact. Besides, compliance with the RD requirements for safety and special TGs and OSTs for the turbine plant shall be demonstrated.

6.2 - 6.11. Third circuit systems

6.2. Main steam pipeline system

6.3. Feedwater pipeline system

6.3. Feeding and deaeration plant.

6.4. Firing system

6.5. Auxiliary steam system

6.6. Deaerator makeup system

6.7. Main condensate system

6.8. Condensate purification system

6.9. Condenser vacuuming system

6.10. Condenser water pipeline system

6.11. Heating plant

Information on the systems included into the third circuit shall be presented in accordance with the standard structure described in the section "General requirements"; in this case impact of each system on the turbine plant operation and dependent failures affecting the RP shall be specified.

Presentation of the information in subsections 6.2-6.11 shall be focused on the system reliability issues.

Conclusion on the impact of any failures in each of these systems on the SS actuation and maintenance of heat removal from the RP via the third circuit shall be made.

Information on any potential missiles in case of any ruptures of high-pressure pipelines or pressurized vessels in each system shall be provided, and any possible impact of such ruptures on the SS functioning and maintenance of heat removal from the RP via the third circuit shall be demonstrated.

7. REQUIREMENTS FOR THE SECTION "PROCESS CONTROL"

Requirements for the process control system except for the requirements for the CSS described in Section 12 shall be presented in this section.

7.1. Process control system for the NPP power unit

The requirements for the following facilities shall be stated:

- the MCR;

- the ECR;

- the information display systems;

- the radiation situation monitoring systems and means;

- control of fire extinguishing systems;

- the PPS;

- the communication and annunciation system and means;

- the diagnostic system.


Information on the conditions and restrictions used as the basis for the CS design, sources of these conditions and restrictions, purpose of the sub-systems, systems and means, safety principles and criteria underlying the system design shall be provided.

7.1.2. Description

Information containing description of the CS, data on its configuration, basic technical characteristics, description of the system operation principle under normal operation conditions and in case of any operational occurrences with due regard for interaction with other systems and devices as well as any associated equipment shall be provided.

Information on the CS hardware and components included into it particularly the systems and means ensuring remote, automated and (or) automatic control of the normal operation systems of the NPP power unit shall be specified. Information on the means ensuring monitoring and presentation of information on the parameters characterizing the RP operation in all possible variation ranges of normal operation conditions, as well as information on any changes of normal operation conditions; information support systems for the operator including the system for in-process presentation of consolidated information on the current state of the RP and the NPP unit safety to the workers shall be provided. Information on the group communication means between the MCR, the ECR and the NPP power unit workers (personnel); individual communication means between the MCR, the ECR and the workers (personnel) ensuring collection, processing, documenting and storage of the information; diagnostics of state and operation modes; diagnostics of the CS hardware and software, the radiation situation monitoring systems shall be presented.

Information on the CS components and means shall also contain the data on their configuration, basic technical characteristics, location, schemes of the systems and hardware, description of their operation principles: under normal operation conditions, in case of any operational occurrences and accidents with due regard for interaction of the systems and devices and the associated equipment.

The initial design information used for safety analysis particularly the methods for assessment and control of the reliability parameters at different stages of the system development and operation shall be provided.

The analysis results for the character and impact of the control and monitoring system failures not representing initiating events of any accidents, analysis and nature of accidents demonstrating the degree of the CS compliance with the design criteria and the requirements of the safety rules and regulations shall be provided.

The information presented in this subsection shall contain the analysis results for reactions of the systems and devices on any external and internal impacts (fires, flooding, electromagnetic interference, short circuits of the primary power supply network, etc.), reactions of the systems on any potential failures and malfunctions (loss of the insulation quality, voltage drops and induced noise, spurious actuations, loss of control, etc.), results of the quantitative reliability analysis, results of the stability analysis for the control and adjustment circuits and their safety impact.

Information on power supply and grounding, protection against external impacts, any systems ensuring maintenance of the conditions in the rooms for the CS hardware shall be presented.

Special attention shall be paid to substantiation of the solutions related to diagnostics and regular state monitoring for the CS and its constituent parts, devices and components, their periodic inspections and functional tests, recording and documenting of malfunctions and failures as well as training of the workers.

Figures, schemes, diagrams, curves and tables required to substantiate the adopted constructive and technological solutions for safety assurance, the information flow schemes and the coding system shall be also provided and described.

7.1.3. Maintenance

Implemented measures and procedures aimed to eliminate any malfunctions and defects in the course of maintenance shall be substantiated.

7.1.4. Commissioning works

Special attention shall be paid to the methods of operability checks for the control and monitoring systems and hardware, their integrated setup, diagnostics and documenting of their characteristics, acceptance criteria and their substantiation.

Information on the commissioning works in comparison with the similar administrative and technical solutions for commissioning of the CS and its components with due regard for trial and testing of counterparts and prototypes shall be provided.

7.2. RP control systems - normal operation control systems

Information on the RP control systems listed in Subsection 4.2, Section 4 shall be provided.

The complete list of the RP CSs and description of each system in accordance with the section "General requirements" shall be presented.

7.3. Main control room


Information on the conditions and restrictions used as the basis for the MCR design, sources of these conditions and restrictions, purpose of the MCR, systems and means, safety principles and criteria underlying the MCR design shall be provided.

7.3.2. Description

Description of the MCR (including the sub-system displaying the position of the CPS control rods with sensors, communication channels and their redundancy) enabling the operator to use information for performance of any actions required to ensure safety shall be provides, as well as the following data:

- the MCR plan;

- general layout of the MCR;

- configuration of the MCR in-process control panels with the control and monitoring devices installed on them;

- general layouts of the MCR consoles and boards with any control and monitoring devices located on them;

- information on location of safety-related control and monitoring devices and information necessary to substantiate ergonomic requirements for their application, arrangement of information and body fields on the control room panels and boards of the control station (stations).

Special attention shall be paid to information on substantiation of the technical solutions including:

- recording of any actions of the workers in case of accidents;

- automatic provision of information on the state of process equipment and automation devices to the operator;

- independent operability check for the process equipment and automation devices performed by the operator in the course of functioning;

- the list of functions performed automatically with submittal of the relevant information to the operator;

- the list of functions performed by the operator. Information substantiating duplication of automatically implemented functions with functions performed with involvement of the operator shall be presented.

It should be demonstrated how the MCR ensures control of the RP and monitoring of the RP and other power unit systems, particularly safety systems in the course of normal operation and in case of accidents.

The operation principle of the MCR and its components in coordination with other systems and any associated equipment under normal operation conditions as well as in case of any operational occurrences including accidents shall be described.

The instrumentation enabling the operator to use information for performance of any actions required to ensure safety shall be described.

Information on substantiation of the workspace sufficiency for all operating personnel both under normal power unit operation conditions and in case of any abnormal operation including accidents shall be provided.

It shall be demonstrated that the CS is designed in such a way so that to provide arrangements for restriction of access to the control rooms and especially to the in-process control zones for the persons not included into the shifts, both in the course of normal operation and in case of any emergency situations.

Information on ergonomic and anthropometric arrangement of the workplaces for operators shall be provided.

The following shall be substantiated for the information fields of the operator's workplace:

- location of the display means for safety-related information on the MCR panels and boards of the control station (stations);

- distinctive coloring of the safety-related information display means;

- convenience of the operator's observation over the displayed safety-related information (zones of vision, size of scales, figures and other symbols);

- reliability of the applied lighting for scales, figures and other symbols on the display equipment.

The following shall be substantiated for the body fields of the operator's workplace:

- location of the controls (buttons, switches, etc.) for safety-related actuators on the body fields of the control room panels and boards of the station (stations) with due regard for convenience of observation over the displayed information required for control through the use of these means;

- distinctive coloring of the controls for safety-related actuators;

- location of the devices for authorized access to the controls of safety-related actuators in case any such requirements are prescribed;

- distinctive configuration of safety-related information means.

The following shall be substantiated:

- lighting intensity at the workplaces of the operators, color, sound and other distinctive characteristics of alarms that should be well identified by the operator and shall have uniform interpretation in all control rooms of the NPP power unit;

- application of the communication means;

- application of the CCTV means;

- application of the MCR information means intended for use by all operators in the shift;

- ergonomics of the technical solutions for manual and automated information recording by the operator at the workplace;

- structural solutions for storage of any documentation necessary for in-process application at the operator's workplace;

- organization of the operator's meals at the workplace in any regular and contingency situations as well as in case of accidents.

7.3.3. Maintenance


Completeness and scope of the metrological support for the devices of the MCR, its constituent parts and components shall be substantiated.




7.3.5. Safety analysis

Reliability analysis results for all components and constituent parts of the MCR shall be provided, and selection of the parameters to be displayed for the operator under normal operation conditions and in case of any operational occurrences including accidents shall be justified. It shall be demonstrated that the selected and displayed parameters ensure submittal of unambiguous and reliable information on compliance with the NPP power unit safe operation limits and conditions to the operator as well as identification and diagnostics of the SS actuation and functioning.

Analysis results for impact of the MCR habitability and survivability supporting systems on its reliability and operability shall be provided.

Results of the analysis proving elimination of common cause failures for the MCR and the ECR shall be presented.

Analysis shall be provided to demonstrate that the operator has sufficient information on performance of any manual operations required for safety (for example, optimal location of the controls, manual operations for the safety hardware servicing, potential unexpected post-accident actions and state monitoring for the safety hardware) and sufficient time to make correct decisions and to perform any actions in case they are necessary.

Information enabling to define that the operator has the possibility to read the data and indications of instruments for monitoring of the conditions in the reactor, the coolant circulation system, the rooms and safety assurance process systems in all RP operation modes including expected operational states and emergency modes shall be provided.

Information shall include design criteria, types of reading units, the number of reading channels, the measurement range for the parameters in these channels, accuracy and location of the instruments.

7.4. Emergency control room


Information on the conditions and restrictions used as the basis for the ECR design, sources of these conditions and restrictions, purpose of the ECR, systems and means, safety principles and criteria underlying the ECR design shall be provided.

7.4.2. Description


Special attention shall be paid to the information demonstrating that the adopted solutions ensure reliable bringing of the reactor into sub-critical state and long-term maintenance of this state, actuation of safety systems and obtaining of the information on the reactor state through the use of the ECR.

Independence of the ECR from the MCR shall be substantiated by detailed description of the adopted arrangements and technical solutions.

The following shall be also described:

- the ECR structure;

- the ECR layout;

- safety-related functions implemented in the ECR;

- configuration of the ECR panels with the control and monitoring devices installed on them;

- the ECR station;

- the ECR station boards with the control and monitoring devices installed on them.

Information on location of safety-related control and monitoring devices and information necessary to substantiate ergonomic requirements for their application (arrangement of information and body fields on the control room panels and boards of the control station for the operator) shall be provided.

7.4.3. Maintenance


Completeness and scope of the metrological support for the devices of the ECR, its constituent parts and components shall be substantiated.

Special attention shall be paid to the information justifying the adopted solutions for the regulations on the ECR operability maintenence in the course of normal operation.





The list of safety functions implemented in the ECR as well as the information required to substantiate impossibility of any common cause failure of the MCR and the ECR and conditions for the MCR operating personnel moving to the ECR in case of any MCR failure shall be provided.

Solutions for assurance of the ECR habitability and survivability in case of design basis and beyond design basis accidents shall be analyzed.

7.5. Information display systems

7.5.1. Description of the information display systems

The following shall be specified in the description of each information display system:

- structure of the system automation means;

- functions performed by the system automatically;

- functions performed by the operator in accordance with the information obtained from the system;

- description of the control of supporting systems required for operation of the information display system;

- the information display system.

Safety requirements the design of each information display system shall comply with should be specified.

The subsection shall include the design materials and any other materials required to substantiate safety and compliance with the RD requirements.


- the system operation algorithms;

- characteristics of the system channels with indication of implementation quality for each function;

- design information on location of the system automation means;

- technical capabilities for implementation of the system functions in case of design basis and beyond design basis accidents when any failures of the system and its supporting devices are possible;

- description of diagnostics for the information display channels;

- instructions on the operator's involvement in the system functioning and control of the activities.

The materials of this subsection shall confirm that the operator has sufficient information to perform any distant manual operations required for safety (for example, changes on the position of controlled rods in the nuclear core; operability check for the monitoring channels of safety-related paremeters; power recording, etc.) and sufficient time to make and implement the decisions.

Data enabling to define availability of the following information to the operator in all RP operation modes shall be specified:

- on the parameters determining the reactor state;

- on the parameters of the coolant circulation and heat removal system;

- on the parameters determining conditions in the reactor containment and the process safety systems including their control automation devices.

Documents with the description of the display systems shall contain the information for the operator concerning implementation quality of all functions for display of safety-related information (reliability characteristics, metrological characteristics, the required stability characteristics, for example under the impact of electromagnetic interference, humidity, etc.), and documents for each information display system developed and accepted for operation containing the data required to substantiate the system compliance with the established requirements.

7.5.2. Analysis of compliance of the systems for information display to the operator with the safety requirements

The materials of this section shall contain substantiation of the way to assure compliance with the prescribed safety requirements, particularly:

- the system functional reliability analysis results;

- analysis results for the consequences of its failures.

7.6. Radiation situation monitoring systems and means

7.6.1. Description of the systems

The following shall be specified in the description of each system:

- structure of the system automation means;

- functions performed by the system automatically;

- functions performed by the operator within the system;

- description of the supporting systems.

7.6.2. Safety requirements for the design of each system

The design materials and any other materials required to substantiate compliance with the RD requirements shall be provided.

Particularly, the following shall be specified for each system:

- the system operation algorithms;

- characteristics of the system channels with indication of implementation quality for each function;

- design documentation for location of the system automation means;

- technical capabilities for implementation of the system functions in case of design basis and beyond design basis accidents when any failures of the system and its supporting devices are possible;

- instructions on the operator's involvement in the system functioning and control of the activities.

7.6.3. Documents for each system developed and accepted for operation containing the information to substantiate compliance of the system with the established requirements

The list of documents for each system developed and accepted for operation containing the information to substantiate compliance of the system with the established requirements shall be provided.

7.6.4. Analysis of the system compliance with safety requirements

The way to assure compliance with the prescribed safety requirements shall be substantiated, including:

- the system functional reliability analysis results;

- analysis results for the consequences of any system failures.

7.7. Control systems of the fire extinguishing systems

Description of the fire extinguishing system and information on control of the fire extinguishing systems not classified as supporting safety systems shall be provided.

7.8. Communication and annunciation systems and means

Description of the system similar to the description given in Subsection 7.6 shall be provided.

Description of the systems and means for preventive and emergency warning of the NPP power unit workers shall also contain:

- the list of warning signals with indication of any light, sound and other signals accompanying them in order to attract attention of the workers;

- technical characteristics of the means intended to attract attention of the workers (flashing frequency, color, pitch of tone, etc.).

Information on the adopted preventive and emergency warning system for the NPP power unit workers shall contain the rules for usage of the warning signal system in emergency situations.

Information on the communication means, particularly redundant ones, intended for the NPP control arrangement as well as the annunciation systems under normal operation conditions, in case of any design basis and beyond design basis accidents shall be provided.

7.9. Diagnostic systems

Description of the diagnostic systems including the diagnostic systems for safety barriers and usage of the diagnostics results (signals) for automatic and automated control shall be provided.

8. REQUIREMENTS FOR THE SECTION "POWER SUPPLY"

Information confirming functional development and reliability of the power supply systems, capacity sufficiency, multi-channel arrangement, independence, resistance to external and internal impacts, the possibility for maintenance, testing and repair, compliance with the requirements of the safety standards and regulations based on the functional analysis of the systems under normal operation conditions and in case of any operational occurrences and failures of power supply systems with due regard for human errors as well as during design basis and beyond design basis accidents shall be provided. Qualitative and quantitative reliability analysis for the power supply of safety systems shall be presented.

The basic principles of the design and operation arrangement for the electrical systems of the NPP power unit shall be provided.

Brief description of the technical solutions accompanied with graphical information reflecting structural peculiarities of the system shall be presented.

8.1. External power grid

8.1.1. Power output diagram


- Development of the power grid

- The NPP purpose and function in the regional power grid

- Characteristics of the power output diagram and the main electric circuit diagram

- The possibility for the NPP power output to the regional substations

- Protection of the networks and substations against external impacts

- Emergency protection automatics, its structural scheme and quantitative reliability characteristics

- Overvoltage protection

- Voltage variations

- Automated dispatch control system

- Operation arrangement for power networks

- Requirements for the NPP cyclic load capability

8.1.2. Characteristics of the power grid


- short circuit current in the NPP power unit circuits

- the possibility to provide auxiliary power supply of the unit;

- sufficiency of the demand management capacity of the system for operation of the power unit in the basic mode, possibility to limit the power of any other generation sources except for the NPP as well as the description of the cases when the necessity to limit the NPP unit power may occur in the grid;

- the possibility to control the system frequency in manual and automatic modes in case of any system accidents;

- the possibility for automatic or manual isolation of the NPP power unit from the power grid with switching to the auxiliary power supply mode;

- permissible unit power of a single NPP power unit on the conditions of the power grid stability maintenance in case of its automatic or manual outage;

- the possibility for the NPP power unit separation for balanced load in case of any system accidents;

- types and intensity of disturbances in the power grid operation;

- the capability for the complete NPP unit power output in case of any disturbances in the grid;

- type of the turbine generator excitation system on the conditions of the power grid stability maintenance;

- the possibility to arrange power supply from the auxiliary NPP power supply system in case of any natural external impacts (earthquake, hurricane, glaze ice, atmospheric pollution, etc.);

- impact of the power supply system on the unit operation;

- impact of the power supply system operation and the main electric circuit by the types, frequency and duration of malfunctions including complete blackout of switchgear;

- analysis of impact of various malfunctions on the NPP power unit safety.

The following types of malfunctions shall be considered:

- total blackout of the power unit due to loss of connections to the power grid;

- frequency deviations;

- three-, double- and single-phase short circuits;

- voltage variations;

- synchronous and asynchronous swings in the power grid, particularly asynchronous swings caused by control failure.

8.2. Main electric circuit diagram

8.2.1. General description

Compliance with the RD requirements shall be demonstrated, the scheme for the generator connection to the mains in order to ensure the maximum achievable reliability of the auxiliary power supply for the unit and the primary switching scheme shall be substantiated.

The fire safety assurance means shall be listed.

Protection schemes and setpoints for the main circuit equipment shall be provided.

8.2.2. Power generator, unit transformer and their auxiliary systems

General description and technical characteristics of the main and auxiliary equipment shall be provided:

- power generator;

- unit transformer;

- electrical primary switching schemes;

- fire and explosion protection;

- substantiation for any deviations from the RD requirements;

- schemes of relay protection devices and emergency protection automatics.

8.2.3. Main circuit control rooms

Description of the main circuit control rooms with measurement and alarm systems shall be presented. Their operability under internal and external impacts shall be substantiated.

8.3. Auxiliary system

8.3.1. Auxiliary power supply system

8.3.1.1. AC/DC auxiliary power supply

The auxiliary power supply system diagram shall be presented.

Description of the operating and redundant power supply sources located at the NPP power unit site and outside it as well as quantitative assessment of their reliability shall be provided. Independence of the power supply sources for the consumers ensuring integrity of the main equipment, fire safety and the NPP power unit start-up and shutdown shall be demonstrated.

Technical specifications of the equipment, hardware, cables, buses, insulators, etc. shall be specified. Their compliance with the RD requirements shall be demonstrated.

The primary switching schemes shall be provided.

8.3.1.2. Short circuit current calculations

Results of calculations for selection of electrical equipment, hardware, buses, insulators and cables, calculations of the protection parameters and parameters of automatic devices, the possibility for self-starting of the auxiliary NPP power unit consumers as well as diagrams of protection, automation and other secondary switching circuits and short circuit current values shall be presented.

8.3.1.3. Selection of actuation setpoints

Selection of the actuation setpoints for the ASB and automatic devices for switching of the reliable power supply network to independent power supply as well as auxiliary operation of the generator in the rundown mode with the frequency and voltage parameters below the permissible limits shall be substantiated.

8.3.1.4. Layout plans for the equipment, hardware and cables

Layout plans shall be provided for the equipment, hardware and cables.

8.3.1.5. Overvoltage protection

Any potential overvoltage and the relevant protection shall be specified.

8.3.1.6. Fire safety assurance

Information on fire safety assurance including description of the automatic fire detection and extinguishing systems with the results of the relevant calculations shall be presented.

8.3.1.7. Causes of fires

Any possible causes of fire breakouts in the electrical section of the NPP, fire propagation paths and their impacts on safety shall be analyzed.

8.3.1.8. Equipment protection

Protection of the electrical equipment of the NPP power unit against any unintended erroneous actions of the personnel (impossibility to activate any equipment with disabled protections and interlocks; availability of automatic devices aimed to change the logic of protections and interlocks in case of any individual equipment withdrawal from service; automatic control of correct assembly of electric and process circuits; impossibility to disable protections and interlocks without the relevant automatic change of operation modes for the main and auxiliary equipment) shall be confirmed.

8.3.1.9. Monitoring and control

Information on the control rooms, the controlled parameters, alarm types, classes of instruments, sensors, measuring transformers, metrological control and protection against external and internal interference shall be provided.

Information on the sensors used in the CSSs shall be specified.

8.3.1.10. Reliability of power supply

Results of the qualitative and quantitative reliability analysis for auxiliary power supply of the NPP shall be provided.

8.3.2. Fire protection of cable systems

8.3.2.1. Applied types of cables


- conditions of flammability, fire resistance, flame retardance, smoke emission and toxicity;

- conditions for flame retardance of separate cables and cable bundles.

8.3.2.2. Cable laying techniques in the areas with various hazards

The cable laying areas shall be characterized by the explosion hazard amd mechanical damages in case of a fire.

8.3.2.3. Passive protection methods


- fire-resistant envelope structures;

- fire partitions restricting the fire propagation through walls and floors and along the cable routes;

- fire-resistant coatings and other arrangements aimed to reduce fire hazard of the cable routes laid in the same fire zone.

8.3.2.4. Active protection methods


- fire alarm;

- automatic fire extinguishing;

- assurance of the maximum ambient air temperatures in normal and emergency modes including blackout.

8.3.2.5. Overheating protection in case of overloads

Proofs of heat and fire resistance in case of any overloads and short circuits shall be provided.

8.3.2.6. Protection against external and internal impacts

Technical solutions for protection against external and internal impacts shall be described.

8.4. Emergency power supply system

Requirements for the electrical section of the EPSS classified as a safety system are given in Section 8 (as the part supporting the EPSS SS).

The EPSS CSS including the step-by-step start-up (loading) programs is only mentioned in these sections, and its complete description is given in Section 7.


Purpose, design basis and classification of the emergency power supply system as a safety system shall be specified.

8.4.2. Description of the system and equipment

8.4.2.1. Characteristics of loads

The list of auxiliary consumers requiring power supply from independent sources in case of any loss of power supply from normal operation sources shall be provided with indication of the permissible power supply interruption duration for each consumer.

8.4.2.2. Technical characteristics of the emergency power supply system


- configuration of the system;

- the electric primary switching diagram of the system with substantiation of its selection;

- sufficiency substantiation for the selected number of EPSS channels;

- sufficiency substantiation for the continuous functioning time of power supply sources;

- technical characteristics of the current sources, power (rated and maximum), permissible duration of continuous operation, voltage and frequency stability, any potential deviations of the current curve from sine wave, the process diagram of the SDGS;

- nameplate data or technical characteristics of the equipment, buses, insulators, cables, hardware, penetration seals, etc. applied in the system, as well as description of the algorithm for switching to independent power supply sources;

- calculation results for short circuit currents and line-to-earth fault currents, selection of the electrical equipment, hardware, buses, insulators and cables;

- potential overvoltage levels and the relevant protection;

- substantiation of the neutral point connection selection (grounded or non-grounded) in order to ensure the maximum reliability of power supply for essential consumers;

- confirmation of the system protection against unintended erroneous actions of the personnel in the course of its actuation (impossibility of switch-on without activation of the relevant protections and automatic devices, automatic control of correct assembly of electric and process circuits, etc.);

- buildings and facilities where the system is located;

- layout plans of the EPSS equipment, hardware and cables as well as electric drives, their switching equipment and the SS cables;

- fire protection substantiation with the calculation results for the maximum temperatures that the envelope and load-bearing structures can reach in case of complete combustion of flammable substances in a single cable room or separate equipment box; results of calculations confirming sufficient strength of these structures and impossibility of further fire propagation, particularly due to heat transfer along the cables.


The EPSS control system is classified as a CSS.


- control rooms, their survivability in emergency situations and under external impacts;

- the controlled parameters;

- types of alarm;

- classes of instruments, sensors, measuring transformers;

- types of protection, their purpose and areas of coverage, technical characteristics, priority of protections;

- redundancy rate for the protections, the domination principle;

- protection against external and internal interference;

- protection against arc discharges;

- calculation for selection of protections and their setpoints;

- requirements for reliability of internal protections for the electrical equipment, cables and diesels with indication of their activation priorities in relation to performance of the safety functions by this power supply system;

- selection of setpoints for the automatic devices (ASB, etc.) and their substantiation;

- protection schemes for the devices, automatics and other secondary switching circuits;

- metrological control;

- protection against short circuits and line-to-earth faults.

8.4.2.4. Testing and maintenance

Information on the following issues shall be provided:

- continuous automatic diagnostic control of the systems and components;

- frequency of testing, testing methods and programs;

- conditions of testing (on operating or shut-down equipment):

- types and time limits for maintenance of the equipment, switching devices, protection and automation cables;

- time limits for replacement of the equipment and cables with exhausted lifetime;

- accessibility for maintenance and testing with due regard for the radiation hazard and ambient conditions.

8.4.2.5. Criteria for selection of the power supply source capacity

Information on the following issues shall be provided:

- calculations of loads on transformers, diesel-generators, power supply lines, inverters and batteries, charging devices;

- coordination of the source capacity with design loads;

- coordination of the load characteristics (active, reactive, inductive) with characteristics of the sources;

- permissible voltage and frequency variations, deviations from harmonicity, inrush currents and asynchronous ASB currents;

- characteristics of batteries with confirmation of their compliance with the requirements of the consumers;

- substantiation of the battery operation time in autonomous mode without any charging;

- characteristics of the charging devices;

- electromagnetic compatibility of the sources, consumers, protections and automatic devices.

8.4.2.6. Location, protective grounding, lightning protection, fire protection

Physical separation of the rooms for switchgear, power supply sources and cable routes as well as their protection against external impacts (earthquake, ASW, aircraft crash, dust storms, salty fogs, chemical and radiation pollution of the atmosphere) shall be demonstrated.

The following shall be briefly described:

- lightning protection and protection against secondary lightning effects;

- protective grounding;

- fire alarm and fire extinguishing systems;

- protection of the equipment, cables and leak-tight penetrations against any missiles in case of any breakage of process equipment and pipelines and hydrodynamic impacts in case of accidents.

8.4.2.7. Conditions for selection of the equipment, cables and penetrations

Conditions for selection of the equipment, cables and penetrations shall be provided, namely:

- ambient conditions;

- seismic resistance;

- power and load-carrying capacity;

- equipment resistance to short circuit currents, heat resistance of cables, particularly heat resistance in case of short circuit current tripping by stand-by protections and after repeated voltage supply with non-eliminated short circuit;

- dust and moisture protection;

- start and self-start assurance;

- thermal class of insulation;

- insulation class with regard to contamination;

- service life, possibility for recovery and replacement;

- resistance to external and internal impacts;

- fire safety.


Consultant Plus: note.There is probably a misprint in the official document test: there is no Subsection 5 in the section "General requirements".

Analysis of the design within the scope of the requirements stipulated in Subsection 5, the section "General requirements" of this regulatory document shall be provided in accordance with the following sequence:

- the system reliability;

- normal operation;

- functioning in case of any failures;

- functioning in case of accidents and external impacts;

- safety analysis;

- compliance with the RD requirements.

8.5. Power supply in case of beyond design basis accidents

Information on the unit power supply in case of beyond design basis accidents shall be provided.

8.6. Operation

References to the information specified in Section 14 may be given.

8.6.1. Operation manuals

General provisions of the power supply system operation manuals shall be presented including:

- the procedure for performance of works and switchings in order to activate individual equipment and systems and to take them out of service for repair;

- the procedure for trial of individual equipment and systems in general;

- frequency of trials;

- quality control for fuel and oil, time limits, criteria and procedure for their replacement (referred to the SDGS);

- frequency and procedure for inspections of the equipment and rooms of the systems.

8.6.2. Repair instructions


- scope and frequency of the equipment repair, checks of protections and automatics;

- time limits and procedure for replacement of the equipment with exhausted lifetime;

- frequency and scope of inspections for the measurement equipment.

8.6.3. Commissioning

The adjustment, trial and testing programs for individual equipment, hardware and the systems in general including the scope of protection and automation checks shall be provided.

8.7. Communication

The basic principles of the communication arrangement shall be specified. Similar information shall be provided in Section 9.

8.8. Standards and regulations

The list of mandatory standards and regulations used in design of the electrical systems shall be presented in tabular format with indication of the systems they are applicable to.

9. REQUIREMENTS FOR THE SECTION "AUXILIARY SYSTEMS"

9.1. Set of nuclear fuel storage and handling systems

The configuration of the set shall be presented in the subsection introduction including the following systems:

1. Fresh (non-irradiated) NF storage and handling system

2. Nuclear core refueling system

3. SNF handling system consisting of:

- the ex-core SNF storage system;

- the system for SNF storage in the repository located outside the reactor hall in the spent fuel pool specially constructed for this purpose;

- the SNF washing system;

- the shielding chamber (if any).

Issues related to NF transportation within the NPP territory from acceptance of a vehicle with fresh nuclear fuel to acceptance (shipment) of spent nuclear fuel shall be specified.

Arrangement of the NF accounting and control at the NPP power unit shall be described.

9.1.1. Fresh (non-irradiated) nuclear fuel storage and handling system

9.1.1.1. Purpose and classification

Information on the purpose of the system with indication of all performed functions shall be provided.

Class, category, safety and seismic resistance group shall be specified for the fresh NF storage

system and its components in the FFS in accordance with the classification used in the effective regulatory documents.

The list of safety RDs containing the requirements the described system shall comply with shall be provided.


The main principles and criteria taken as the basis for the system design shall be provided.

The following shall be specified for each repository:

- the maximum design capacity of the repository;

- storage norms;

- characteristics of the fresh fuel planned for storage (enrichment, dimensions, activity level, heat emission level, etc.);

- distinctive signs characterizing fuel enrichment in fuel assemblies and techniques for their identification - visual and (or) through the use of refueling devices;

- distinctive signs for fuel assemblies with burnable poison, combined fuel, including uranium and plutonium one, etc. (if any) and techniques for their identification.

Lists of methodologies and programs used to substantiate safety of NF storage and transportation shall be presented, their application scope as well as information on verification and validation of the methodologies and programs in accordance with the established procedures shall be specified.

The list of parameters, sub-systems, system components defined in the design in order to ensure its safe functioning shall be provided.

The list of design initiating events the system is designed for shall be specified. Load combinations for the calculation shall be presented.

Special requirements for the systems associated with the main system functioning shall be stated.

The main principles and criteria taken as the basis for the system layout solutions shall be provided.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type and necessity for its storage the possibility to use the existing fresh NF storage facilities for this purpose shall be confirmed, or the design materials for modification of the fresh fuel storage facilities as well as any possible modification of the handling equipment components shall be provided.


Description of the design and (or) process flow diagram of the system in general, its sub-systems, equipment, structures and components (if they perform independent functions) shall be presented.

Drawings, figures and schemes illustrating the design and operation of the system and its components, its spatial layout and interfaces with other NPP power unit systems shall be provided.

1. Description of the system location

The internal layout of the repository shall be described, the repository class and the storage environment parameters (temperature, humidity, etc.) as well as the RD requirements for safety shall be

specified. It shall be particularly demonstrated that the layout of the rooms and design solutions eliminate any possibility of flooding with water and ingress of any other neutron-moderating materials into the non-irradiated fuel storage areas; fast evacuation of the personnel from the rooms is ensured in case of any accidents (accident type, evacuation routes, evacuation time calculations); no routes to any other operational rooms pass through the fuel storage facilities (the system of access and access control shall be described).

The repository layout in the building shall be described with indication of its location in relation to any other rooms of the NPP power unit, the plant and the adjacent systems.

The following shall be specified (in the absence of information in Section 2):

- classification of the FFS building and facilities (if any) in accordance with safety and seismic resistance categories;

- techniques and methods to comply with the prohibition for transportation of any cargoes (except for the parts of hoisting and refueling devices) above the stored fuel in the course of refueling or location of any cargoes above the storage facilities covered with any structures; confirmation of the capability of these structures to withstand dynamic and static loads occurring in the course of cargo transportation or location;

- information on division of the FFS buildings and rooms into the controlled access areas and the uncontrolled access areas;

- information on classification of the FFS rooms into categories according to radiation and fire safety and information on the FFS rooms where the radiation situation can charge drastically in the course of process operations;

- information on compliance with the principle for separate ventilation in the FFS rooms of the controlled access area and the uncontrolled access area as well as absence of any common ventilation air ducts in the rooms with different service categories;

- information on installation of leak-tight doors at all emergency (fire) entrances and exits from the controlled access areas;

- information confirming that the design of the repository (in case of necessity) enables to decontaminate any surfaces easily, and the surfaces in the controlled access area rooms are protected with materials with poor RSb absorption that are easy to decontaminate.

2. Description of the fresh NF storage system equipment:

- configuration of the fuel storage and handling system equipment shall be specified, brief description of its design shall be presented including the equipment used for fuel storage, handling and canting operations, depreservation, inspection (incoming control) and repair of fuel assemblies (if any);

- the service systems for the transport packings (if any are provided in the FFS) shall be described.

In case of any reactor core refurbishment associated with usage of new fuel type and necessity for its storage in ISTPs (TPSs) in the FFS applicability of the existing ISTSs (TPSs) for this purpose shall be confirmed, or materials of the new ISTP (TPS) design ensuring non-exceedance of the standard radiation burden values on its surface shall be provided, and information on the radiological monitoring measures and maintenance conditions for transport packings with such fuel shall be specified.

3. Information on any other equipment and materials stored in the FFS


- techniques and methods used to comply with the prohibition for storage of any flammable materials as well as materials with other hazardous properties in case of a fire not included into the transport packings in the FFS;

- in case any other nuclear core components except for the nuclear fuel are stored in the FFS - the list of these components and design regulations for their locations;

- techniques and methods to comply with the prohibition for storage of any efficient neutron-moderating materials between or inside casings, between the racks or packaging groups.

4. Information on the systems associated with functioning of the set of fresh NF storage and handling systems shall be provided, and any systems, sub-systems, equipment, structures and components performing independent functions shall be specified:

- information on location of each system, configuration of its equipment, redundancy, specified service life, working media, parameters, etc. The information shall include the parameters corresponding to the functional purpose of the described system. The parameter values shall be specified with indication of potential scattering (with the margin);

- confinement devices intended to prevent or limit propagation of radioactive substances and ionizing radiation generated in case of accidents inside the storage facility and their release to the environment;

- SSCR EAS;

- the fire alarm system;

- the operating and emergency lighting system;

- closed-circuit television (if any);

- ventilation systems;

- drainage systems;

- the communication system;

- the set decontamination system;

- the repository heating system.

In case of any reactor nuclear core refurbishment associated with usage of new fuel type and necessity for its storage in the existing fresh fuel storage facilities sufficiency of the existing systems related to the fresh NF storage system functioning shall be confirmed, or design materials for refurbishment of such systems shall be provided.

9.1.1.4. Materials

The minimum amount of information on the materials shall include:

1. Information on the materials planned for usage in the main system components including welding materials, their mechanical and process characteristics. References to any technical specifications, GOSTs, etc. may be given. The information shall also demonstrate compliance with the requirements for the supply of equipment, instruments, materials and products for the nuclear power facilities. Information confirming compliance with these requirements shall be provided for the FFS handling equipment subject to the requirements of the effective regulatory documents.

2. Information on the permits for application of the above-mentioned materials, including information on permits for application of non-metal materials (if any) in case they are required in accordance with the regulatory documents for safety. In the absence of this requirement the relevant record shall be provided in the section.

3. Special information on the stability of materials, particularly absorbing additives in the FFS structural materials (if any), under the conditions occurring in the course of operation, including decontamination, and in case of any operational occurrences, including accidents; this information shall demonstrate compliance with the RD requirements.

4. Special information demonstrating in particular:

- compliance with the requirements for fire resistance or slow burning of lining, finishing, sound-absorbing, sound- and heat-insulating materials used for interior finishing of the FFS;

- confirmation of the fact that the FFS envelope structures are made of non-combustible materials and have the fire resistance rating in accordance with the requirements;

- protection of the surfaces in the FFS rooms and the FFS equipment with moisture-resistant materials having low absorption of radioactive substances and easy to decontaminate;

- information on any hazardous properties of the materials used and stored in the FFS (if any) in case of any potential manifestations of such properties in the course of normal operation and under any operational occurrences including accidents.

9.1.1.5. Monitoring and control of the system

The list and substantiation of the permissible values for the controlled parameters of the system in all operation modes and in case of shutdown for repair shall be provided; location of the measuring points shall be specified, the control methods shall be described, and information on metrological validation of the applied methods and requirements for the instrumentation shall be presented. Interfaces of the system with control systems of the unit, redundancy of sensors and communication channels shall be described (references to the information presented in Sections 7 and 8 may be given).

The monitoring systems shall be described with indication of the measuring schemes, points and techniques, the controlled parameters, the protection actuation setpoints, accuracy and frequency of measurements, the assessment criteria and methods.

Information on the presence of any monitoring and alarm devices and systems in the FFS shall be provided.

Information on all types of monitoring and alarms shall be presented.


The information on the quality assurance program complying with the RD requirements shall be provided.

9.1.1.7. Commissioning

Information shall be presented in accordance with the requirements of Section 13.


Information on the regulations and procedure for the regular inspection of the FFS equipment and systems in the course of operation shall be provided.

Information on the methods, scope and time limits for state monitoring and testing of the systems in the course of the NPP operation, characteristics of the arrangements provided in the design for these purposes shall be presented, and their compliance with the RD requirements for safety shall be demonstrated.

9.1.1.9. Normal functioning of the system

Functioning of the system under normal operation conditions and interaction with other systems shall be described.

Information on the operational procedures in the fresh NF storage and handling system shall be provided within the scope complying with the requirements of Section 14.

9.1.1.10. Functioning of the system in case of failures

Analysis of failures of any system components including human errors shall be provided, and the impact of failure consequences (particularly common cause failures) on the system operability and safety of the entire NPP power unit shall be assessed.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type the reviewed list of design basis accidents and the list of beyond design basis accidents in the course of fuel handling considered in the design shall be presented with due regard for peculiarities of the new fuel types that shall be analyzed in Chapter 15.

9.1.1.11. System reliability analysis

1. The calculation programs used for the system reliability analysis, the input data for calculations, assumptions and restrictions adopted for the algorithms and calculation schemes, the calculation results and conclusions shall be described. Information on verification and validation of the calculation programs shall be provided.

The scope of information shall be sufficient to perform independent calculations. In case any experiments were performed to substantiate reliability of the system design the experimental unit and the experiment conditions shall be described, their compliance with the design conditions shall be analyzed, metrological support of the experiments and interpretation of the results with regard to the design conditions shall be presented.

The lists of initiating events, failures, external impacts, human errors and their combinations to be taken into consideration in the analysis of accidents in the system and the NPP reliability in Section 15 shall be provided.

2. Quantitative parameters of the FFS equipment reliability in accordance with the technical specifications for manufacturing shall be presented.

Qualitative reliability analysis shall be provided and quantitative values of the reliability parameters shall be determined for the system (the transport and handling scheme of fresh NF acceptance and supply).

Calculation of the quantitative reliability parameters for the system shall be preliminarily accompanied with a brief description of the calculation program including assumptions, restrictions and information on the program verification.

Results of calculations for determination of the quantitative reliability parameters and analysis of the obtained results shall be provided, and the conclusions on their acceptability or unacceptability shall be stated.

The scope of information shall be sufficient to perform independent alternative calculations in case of necessity.

9.1.1.12. Assessment of the fresh NF storage project

Compliance with the RD requirements for safety shall be analyzed.

The conclusions shall be made based on the criterion for satisfactory compliance of the fresh NF storage and handling systems with the safety requirements defined in the design and compliance of the design with the RD requirements.

Compliance with the radiation safety principles stipulated in the regulatory documents on radiation safety shall be assessed.

Techniques and methods for determination of the permissible number of packages or casings in a group or stack shall be described.

9.1.2. Nuclear core refueling system

Requirements for the nuclear core refueling system shall be specified.


Information on the purpose and classification of the nuclear core refueling system components shall be provided.


The information provided in par. 9.1.1.2 shall be used.

9.1.2.3. Description of the refueling system

1. Description of the process flow diagram

The process flow diagram of the refueling operations shall be described with indication of the equipment, devices and components performing independent functions. Configuration of the particular system equipment shall be specified.

The design flow diagram of handling operations in case of unloading of the nuclear core and its components shall be provided, and its distinctions from the refueling pattern as well as any special-purpose equipment shall be specified.

Particularly, the following shall be described (specified):

- techniques and methods used to identify unloaded fuel assemblies and (or) core components for compliance with the refueling plan;

- the selected refueling technique and its substantiation;

- state of the reloading box and SFAD in the course of refueling;

- the system and design of the unit for the core components loading into the reactor;

- refueling frequency, scope and procedure and substantiation thereof;

- engineering features provided in the design in order to prevent accidental ingress of any foreign objects into the reactor in the course of refueling and repair works;

- configuration of the refueling system with its adequacy substantiation and also with indication of the requirements for this system ensuring safe handling of fuel assemblies, particularly in case of any failures and damages;

- engineering features ensuring heat removal from the reloaded fuel assemblies.

Besides, the following shall be described:

- measures to prevent damage, deformation, breakage or falling of fuel assemblies;

- measures to prevent application of impermissible forces to the fuel assemblies in the course of their withdrawal or installation;

- engineering features aimed to prevent falling of fuel assemblies in case of any power supply loss;

- provided protective devices ensuring movement of refueling devices within the permissible limits;

- equipment provided in the basic design for reliable transportation of fuel to safe locations in case of any failure or disturbances of safe operation conditions for the refueling devices;

- technological devices preventing withdrawal of fuel assemblies with high decay heat;

- consoles (panels) provided in the refueling devices in order to display information on the position (state) and orientation of the fuel assemblies and grips.

2. It shall be demonstrated that all loads occurring under normal operation conditions including asymmetric loads and loads caused by acceleration are taken into account in the design of the NF refueling equipment; in this case it shall be demonstrated that any stresses arising due to impact of these loads do not exceed the acceptable limits for various equipment components.

3. Operability of the refueling system shall be substantiated.

4. Information on the systems associated with functioning of the nuclear core refueling systems shall be provided.

Brief information on location of each system, configuration of its equipment, redundancy, expected service life, working media, parameters, etc. shall be provided.

Information on the following systems shall be presented:

- closed circuit television for the refueling monitoring with the list of refueling operations controlled through the use of closed circuit TV;

- leak-tightness monitoring for the claddings; particularly the criteria for definition of damaged fuel shall be specified, and actions performed to detect such damages shall be described;

- fuel burn-up monitoring;

- operating and emergency lighting;

- fire extinguishing;

- ventilation and air purification;

- communication and warning;

- emergency alarm;

- decontamination.

9.1.2.4. Materials

Information on the applied materials shall be presented. Description shall be provided within the scope of par. 9.1.1.4.

9.1.2.5. Monitoring and control of the system

The list of the controlled system parameters in the course of operation and in case of shutdown for repair shall be provided, and the permissible values shall be substantiated; location of the measuring points shall be specified, the control methods shall be described, and information on metrological validation of the applied methods and requirements for the instrumentation shall be presented. The system interfaces with control systems, redundancy of sensors and communication channels shall be described.

Description of protections and interlocks shall be provided.

Operability of all control and monitoring systems for the refueling system shall be substantiated, and their functions shall be specified. References to Sections 7 and 8 may be given in this section.


Information on quality assurance in the NF refueling system shall comply with the requirements stated in Section 17.

9.1.2.7. Commissioning

Information on the NF refueling system commissioning shall comply with the requirements stated in Section 13.


Information on the regulations and procedure for periodic inspection of the NF refueling equipment and systems in the course of operation, methods, scope and time limits for state monitoring and testing of the systems in the course of the NPP power unit operation, characteristics of the arrangements provided in the design for these purposes shall be presented, and their compliance with the RD requirements for safety shall be demonstrated.

9.1.2.9. Safe operation conditions

Safe RP operation conditions in the course of refueling shall be provided.

9.1.2.10. System reliability analysis

Information complying with the requirements stated in par. 9.1.1.11 shall be provided with regard to the NF refueling system.

9.1.3. Set of systems for handling of spent (irradiated) fuel

9.1.3.1. Ex-core SNF storage system

1. Purpose and classification

2. Design basis

Information complying with the requirements stated in par. 9.1.1.2 shall be provided with regard

to the ex-core SNF storage system.

In case of any refurbishment of the reactor nuclear core related to usage of the new fuel type and necessity for this SNF storage the possibility to use the existing SNF storage facilities for this purpose shall be confirmed, or the design materials for modification of the SNF storage facilities as well as any possible modification of the handling equipment components shall be provided.

3. Description of the system


Detailed drawings, figures and schemes illustrating the design and operation of the system and its components, its spatial layout and interfaces with other NPP power unit systems shall be provided.

Descriptions shall be accompanied with the parameters corresponding to their functional purpose.


The maximum design capacity of heat removal in the spent fuel assembly drum, the storage environment parameters (temperature, pressure, etc.) and the SNF storage norms shall be specified for the ex-core SNF storage system. It shall be demonstrated that the SFAD capacity allows holding of nuclear fuel in order to reduce radioactivity and heat emissions and also arrangement of the conditions for unloading of one complete nuclear core at any moment of operation.

In case of ex-core storage of the fuel assemblies with new fuel irradiated in the reactor applicability of the existing spent fuel assembly drums for storage shall be demonstrated, or the design materials for their refurbishment shall be provided.

Characteristics of the fuel planned for storage (burn-up, activity level, heat emission level, etc.) shall be specified.

Information on any other elements (especially fresh NF) on temporary or long-term storage in the ex-core SNF storage facilities shall be provided with indication of the reasons, time limits and norms of storage as well as the properties of these elements.

The layout of SFADs and handling equipment in the building of the NPP power unit shall be described with indication of their location in relation to other rooms of the NPP power unit and the adjacent systems.

The SFAD design and the SNF storage process flow diagram shall be described with indication of the sub-systems, equipment and components performing independent functions.

Support and civil structures of the spent fuel assembly drum shall be described to the extent they affect the safety conditions.

It shall be demonstrated that:

- the SFAD design eliminates any possibility for the coolant loss under normal operation conditions and in case of design basis accident;

- the SFAD design provides for the possibility to detect leakages;

- the possibility for irradiated NF cooling in case of design basis and beyond design basis accidents is ensured.

Design of the equipment used for the SNF location and storage, particularly for leaky fuel

assemblies, as well as the equipment for storage of any other nuclear core components (if any) shall be described.

In case of any new SNF type storage in the existing ex-core storage facilities administrative and technical measures for storage of damaged and leaky fuel assemblies with such fuel shall be provided.

Configuration of the particular equipment of the SNF storage system and compliance of this equipment with the RD requirements for safety shall be specified.

5. Information on any systems related to functioning of the SNF storage and handling system

Information on location of each system, configuration of its equipment, redundancy, expected service life, working media, parameters, etc. shall be provided.

Parameters corresponding to the functional purpose of the described system shall be specified. The parameters values shall be specified with indication of potential scattering (with the margin).


- LSSs intended to prevent or limit propagation of radioactive substances and ionizing radiation generated in case of accidents inside the storage facility and their release to the environment;

- coolant;

- SFAD filling and emptying;

- make-up;

- the intermediate cooling circuit;


- process control;



- emergency alarm.

Functions of the above-mentioned systems shall be specified, and their operability shall be confirmed (references to any other sections containing this confirmation may be given).

In case of any reactor nuclear core refurbishment associated with usage of the new fuel type and necessity for this SNF storage sufficiency of the existing systems related to the SNF storage system functioning shall be confirmed, or design materials for refurbishment of such systems shall be provided.

6. Materials

Requirements for the materials shall be described within the scope of par. 9.1.1.4.

7. Monitoring and control of the system

Requirements for the system control and monitoring shall be described within the scope of par. 9.1.2.5.

8. Testing and inspections

The scope and methods of incoming control, inter-departmental, pre-operational commissioning tests, their metrological support shall be substantiated; the list and permissible values of the controlled parameters and the requirements for the instrumentation used for the purpose of testing shall be presented and substantiated.

9. Quality assurance

The systems, equipment and processes in the repository as well as the civil structures of the system subject to application of the NPP QAP requirements shall be specified.

It shall be demonstrated that the materials, manufacturing methods, supply and storage conditions, etc. comply with the requirements of the design documentation and the RD, and any actual modifications and deviations (if any) including deviations from specific design requirements and regulatory documents shall be substantiated; the documents where these deviations are recorded shall be specified.

Information on the NPP QAP in general shall be provided.

10. Commissioning

Information on commissioning of the entire ex-core NF storage system shall comply with the requirements stated in Section 13.

Results of the SFP testing shall confirm that the SFP lining ensures the prescribed degree of leak-tightness and accommodation of force impacts, etc.

11. Operation

Information on the regulations and procedure for the regular inspection of the ex-core NF storage system equipment in the course of operation shall be provided.

Information on the methods, scope and time limits for state monitoring and testing of the systems in the course of the NPP power unit operation, characteristics of the arrangements provided in the design for these purposes shall be presented, and their compliance with the RD requirements for safety shall be demonstrated.

Information on the operational procedures in the ex-core NF storage system shall be provided within the scope complying with the requirements of Section 14.

12. System reliability analysis

This paragraph shall be presented within the scope of par. 9.1.1.11.

13. Design assessment

Compliance with the requirements, principles and criteria established in the relevant safety RDs shall be analyzed.

Conclusions shall be made based on the criterion of satisfactory NPP compliance with the safety requirements specified in the design and compliance of the design with the RD requirements.

9.1.3.2. System for SNF storage in water or any other cooling medium in the spent fuel pool located outside the reactor hall in the repository (SNFS) specially built for this purpose

1. Purpose and classification

2. Design basis

The paragraph shall be presented in the scope of par. 9.1.1.2 with regard to the spent fuel pool located outside the reactor hall in the repository (SNFS) specially built for this purpose.




Descriptions shall be accompanied with the parameters corresponding to the functional purpose. The parameter values shall be specified with indication of permissible scattering (with the margin).


Description of the process flow diagram shall be presented.

5. Information on any systems associated with the SNFS set functioning


Parameters corresponding to the functional purpose of the described system shall be specified.


- LSSs intended to prevent or limit propagation of radioactive substances and ionizing radiation generated in case of accidents inside the storage facility and their release to the environment;

- water cooling (except for the cases when it is confirmed that exceedance of the design water temperature values in the repository is prevented even without any special-purpose cooling);

- water purification;

- filling and emptying (the drainage system) of the spent fuel pool;

- make-up;

- water supply;

- collection of radioactive water leaks in controlled water discharge sumps (leakage recycling);


- underwater lighting;

- monitoring of the transport container state and decontamination of transport containers (if any are present in the SNFS);

- decontamination of the complex;



- SSCR EAS;

- security alarm;

- power supply.

In case of any reactor nuclear core refurbishment associated with usage of the new fuel type and necessity for storage of such SNF in the spent fuel pool located outside the reactor hall in the existing SNFS possibility for such storage shall be confirmed, or the design materials for modification of the SNFS SFP including the systems related to functioning of the SNF storage system as well as potential modification of the handling equipment components shall be provided.

6. Materials


7. Control and monitoring



Systems, equipment (components) and processes in the repository as well as civil structures of the system subject to application of the NPP QAP shall be specified, and the relevant control and inspection methods or levels shall be defined.

Information on the NPP QAP complying with the RD requirements shall be presented.


The scope and methods of incoming control, inter-departmental, pre-operational commissioning tests, their metrological support shall be substantiated; the list and permissible values of the controlled parameters and the requirements for the instrumentation used for the purpose of testing shall be presented and substantiated.

10. Commissioning

Information on commissioning of the entire SNFS system shall comply with the requirements stated in Section 13.

11. Operation

The paragraph shall be presented within the scope of par. 9.1.1.9 and 9.1.1.10.


The information complying with the requirements of par. 9.1.1.11 shall be provided.

9.1.3.3. SNF washing system

1. Design basis

Information complying with the requirements stated in par. 9.1.1.2 shall be provided with regard to the SNF washing system.




Descriptions shall be accompanied with the parameters corresponding to their functional purpose.



4. Information on any systems associated with the SNF washing system functioning

Information on location of each system, configuration of its equipment, redundancy, specified service life, working media, parameters, etc. shall be provided with indication of the parameters corresponding to the functional purpose of the described system. The parameter values shall be specified with indication of potential scattering (with the margin).


- LSSs intended to prevent or limit propagation of radioactive substances and ionizing radiation generated in case of accidents and their release to the environment;

- cooling;


- process control;

- leak-tightness monitoring for spent fuel assemblies;

- washing media supply;

- radiological monitoring;

- SSCR EAS;

- power supply of the systems and supporting devices;

- CCTV (if any);

- other systems.

Compliance with the RD requirements for safety shall be demonstrated for all systems mentioned above.

5. Materials



The list and substantiation of the permissible values for the controlled parameters of the system in all operation modes and in case of shutdown for repair shall be provided; location of the measuring points shall be specified, the control methods shall be described, and information on metrological validation of the applied methods and requirements for the instrumentation shall be presented.

The system interfaces with control systems of the unit, redundancy of sensors and communication channels shall be described.

The monitoring systems shall be described with indication of the measuring schemes, points and techniques, the controlled parameters, the protection actuation setpoints, accuracy and frequency of measurements, the assessment criteria and methods.


Requirements for in-service inspections and testing of the systems shall be specified.

8. Commissioning

The paragraph shall be presented within the scope of par. 9.1.1.7 and Section 13.

9. Operation

Information on the basic operational procedures in the set of systems for SNF storage and handling shall be presented.




The subsection shall be completed with the analysis of compliance with the RD requirements for safety.

9.1.3.4. Shielding chamber system

1. Design basis

Information complying with the requirements stated in item 9.1.1.2 with regard to the shielding chamber system shall be provided.




Descriptions shall be accompanied with the parameters corresponding to the functional purpose.

In case the shielding chamber is used to handle the new SNF type the design materials for the shielding chamber intended for this purpose or refurbishment of the existing shielding chamber equipment shall be provided.


Description of the process flow diagram shall be presented.

Besides, the following shall be provided:

- information on arrangement of access to the shielding chamber rooms;

- proof of compliance with the NPP SR requirements;

- information on the SNF handling zones in the shielding chamber system where the radiation

situation can change in the course of process operations.

4. Information on any systems associated with the shielding chamber system functioning

Brief information on location of each system, configuration of its equipment, redundancy, specified service life, working media, parameters, etc. shall be provided with indication of the parameters corresponding to the functional purpose of the described system.


- LSSs intended to prevent or limit releases of radioactive substances and ionizing radiation generated in the course of process operations and (or) in case of accidents to the environment;


- lighting (operating and emergency);

- independent radioactive drain system;

- decontamination of the complex;

- gas supply;

- vacuuming;

- power supply of the systems and supporting devices;



- emergency alarm.

Compliance with the RD requirements for safety shall be demonstrated for all systems listed above.

5. Materials



The list and substantiation of the permissible values for the controlled parameters of the system in all operation modes and in case of shutdown for repair shall be provided; location of the measuring points shall be specified, the control methods shall be described, and information on metrological validation of the applied methods and requirements for the instrumentation shall be presented. The system interfaces with control systems, redundancy of sensors and communication channels shall be described.

The monitoring systems shall be described with indication of the measuring schemes, points and techniques, the controlled parameters, the protection actuation setpoints (for example, fire protection), accuracy and frequency of measurements, the assessment criteria and methods.

It shall be confirmed that control and monitoring of the system ensure compliance with the GSR requirements for timely diagnostics of any defects and detection of malfunctions in order to take any measures for their elimination.

All control devices and systems for the shielding chamber shall be specified.


The information on the quality assurance program complying with the RD requirements shall be provided.


The list of periodical in-service inspections and tests shall be presented.

9. Commissioning

Information on the shielding chamber commissioning shall comply with the requirements of Section 13.

Compliance with the RD requirements shall be demonstrated.

10. Operation

The paragraph shall be developed similar to par. 9.1.1.9 and 9.1.1.10 with regard to the shielding chamber system.


Information on the system compliance with the RD requirements for safety shall be provided.

9.1.4. In-plant nuclear fuel transportation system


Purpose and classification of the in-plant NF transportation system shall be specified.


The paragraph shall be presented within the scope of par. 9.1.1.2 with regard to the in-plant NF transportation system at the NPP territory.


Information on the parking area for the transportation vehicle and location of in-plant railway lines for NF transportation, methods and scope of incoming control for containers with nuclear fuel, methods for transfer of unloaded nuclear fuel from the train to the storage facility, the scheme of NF transportation within the NPP site, methods for NF transportation to the power units through the use of in-plant TPSs and special-purpose vehicles shall be presented.

In case of any need for in-plant transportation of new fuel types the possibility for its transportation in the existing ITPSs shall be confirmed, or materials on improvement of the ITPS design for transportation of new fuel types shall be presented, and information on the radiological monitoring measures and transportation conditions for new fuel types by special-purpose vehicles shall be provided.

Any systems associated with functioning of the in-plant NF transportation system shall be described in this subsection from the information integrity considerations to the extent they may be reviewed as the part of this system.

In case the necessary information is presented in any other section or subsection of this document the relevant reference shall be given in this section.



Control and monitoring procedures for the NF transportation shall be described.


Information on in-service control, inspections and testing shall be provided.

9.1.4.6. Operation

Brief description of the basic operational procedures shall be presented.


Information on the system compliance with the RD requirements for safety shall be provided.

9.1.5. Arrangement of nuclear fuel accounting and control

Arrangement of the NF accounting including the NF identification issues (FA type, number, nuclide composition, enrichment, etc.), places of installation (stowage), registration of the time for delivery to the repository and issuance to the RB, maintenance of charts and other accounting documentation shall be described.

Information demonstrating that the accounting and control procedures for fissionable nuclear materials provide reliable information on the quantity and location of nuclear materials, timely detection of any losses and unauthorized use or theft shall be provided, including:

- description of the structure of NM balance areas and key points for measurement of the inventory and flows of nuclear materials with regard to the FFS;

- classification of fissionable nuclear materials into categories;

- description of the recording procedure for any changes in the inventory of fissionable nuclear materials including delivery to and transfer from the MBA with regard to the FFS;

- description of the maintenance of any material balance accounting and operational documents for MBAs and key measurement points, description of the physical NM inventory arrangement;

- description of the procedure for development of reports on MBAs.

9.2. Sodium-contaning auxiliary systems

Information on normal operation of safety-related sodium-containing auxiliary systems shall be presented.

The following systems shall be considered;

- oxide removal from the primary circuit sodium;

- detection of defective fuel assemblies;

- spectrometric control of the primary circuit sodium;

- cesium content monitoring in the primary circuit sodium;

- cesium removal from the primary circuit sodium;

- sodium preparation;

- oxide content monitoring in the primary circuit sodium;

- the primary circuit vessels;

- the primary circuit coolant sampling;

- auxiliary sodium systems of the SFAD.

The above-mentioned list may be adjusted in accordance with the particular design.

It is recommended to follow the description structure specified in the appendix to the section "General requirements" for presentation of the information on any systems.

9.2.1. System for oxide removal from the primary circuit sodium

9.2.1.1. Purpose

Information on the system purpose shall be presented.

The list of the main safety regulatory documents used as the basis for design of this system and containing the requirements it should comply with shall be provided.



- the list of the NPP design modes requiring operation of the system for oxide removal from the primary circuit sodium;

- the input data for design of this system (capacity, pressure, temperature, design modes, purification quality parameters);

- criteria the system shall comply with, i.e. the paragraphs of particular safety regulatory documents, quantitative parameters and safety criteria;

- the list of systems ensuring operability of the primary circuit sodium purification system and performance of the prescribed functions, including any systems for power supply, automatic control and cooling of the system components;

- information on layout with due regard for location of individual systems and components in order to perform the prescribed functions and access to the equipment, taking into account the impacts caused by the coolant leaks and maintenance of the equipment operability in this case, as well as the requirement for location of the components connected to different power supply systems.

9.2.1.3. System design


The process flow diagram and its description shall be provided.

Compliance with the RD requirements for safety and the requirements stipulated in par. 9.2.1.2

shall be demonstrated.

The main characteristics of the equipment shall be specified.


a) Pump set:

- type of the set;

- pressure;

- power;

- temperature of the pumped medium;

- design temperature of the pumped medium;

- sealing of the pump, leakage;

- net positive suction head.

b) Heat exchanger:

- type;

- heat exchange surface area.

c) Tube side:

- medium;

- the pumped medium flow rate;

- design pressure of the pumped medium;

- operating pressure of the pumped medium;

- operating temperature of the pumped medium;

- tube material.

d) Shell side:

- the pumped medium;

- the pumped medium flow rate;

- design pressure of the pumped medium;

- design temperature of the pumped medium;

- operating pressure of the pumped medium;

- operating temperature of the pumped medium;

- material.

Data sufficient to assess safety and operability of the equipment in the entire range of modes shall

be provided for special-purpose equipment (for example, entrainment filters, sodium vapor traps, etc.).

2. Description of the components

Description of the system equipment and its peculiarities shall be presented.

3. Materials

Selection of the materials shall be substantiated with due regard for the following factors:

- properties of the pumped medium and their impact on corrosion of the structural materials;

- parameters of the pumped medium;

- the ambient parameters;

- development of the manufacturing technolofy for the equipment and pipelines.

The basic material applied, techniques and (or) means for protection of the equipment against environmental impacts, the climatic category of the item shall be specified. Information on the materials shall contain references to GOSTs or technical specifications for the material with indication of its mechanical properties and chemical composition. Feasibility of the selected material under normal operation conditions of the system and in case of any operational occurrences including accidents shall be confirmed.

In case any new materials are applied information on validation of the materials and their experimental substantiation shall be provided.

4. Overpressure protection

Means for protection of the systems against overpressure shall be described, and operability of these means shall be substantiated by calculations and (or) experiments.

5. Location of the equipment:


- location of the system equipment in the relevant buildings and rooms and the elevations of its location;

- conditions for location of the components connected to different power supply and control systems;

- fire resistance of the rooms;

- fire safety conditions;

- protection against missiles;

- the systems maintaining the required ambient parameters;

- seismic categories of the relevant buidlings and facilities.

Reference to the layout drawings (plans and profiles) attached to this section shall be given.

6. Electric heating of the system

Information on arrangement of the electric heating system, sequence of its activation, the

requirements for the temperature maintenance and the deactivation sequence shall be provided.


The list of controlled parameters shall be provided.

The list of protections and process interlocks, controllers, control programs shall be presented.

Control of the system by the operator in case of any failure of the automatic system control or any deviations from the normal operation conditions shall be described.

9.2.1.5. Safe operation conditions for the reactor plant

Safe operation conditions for the reactor plant determined by the purification system state shall be specified.


Information on the system testing and inspections including the testing and inspection methodology with indication of the controlled parameters and instrumentation shall be provided.

Frequency of the system testing and inspections shall be specified.

9.2.1.7. Operation of the system

1. Normal operation

Information on operation of the system, its individual components and assemblies in various operating modes of the NPP power unit and performance of the prescribed functions by the system shall be provided:

- cold start-up of the power unit;

- start-up of the power unit after refueling;

- power operation of the unit;

- the system operation mode in the course of refueling;

- shutdown of the power unit with cooldown.

2. Functioning of the system in case of any deviations from the operation limits and conditions

Information on the system functioning and performance of the relevant functions in case of any failures of individual components, the possibility for identification of the relevant system component failures by the operator, the impact of these failures on the system and RP operation as well as on the NPP power unit safety in general shall be provided.

The required information on the operator's actions aimed to confine any malfunction in case of a failure and the method for the RP bringing into safe state shall be provided.

Information on the arrangements aimed to prevent loss of the coolant from the primary circuit and the SFAD cooling system shall be presented.

3. Functioning of the system in case of emergency situations and design basis accidents

Information on operation of the system, alarms, actions of the automatic devices and the operator, the necessity for the RP shutdown and the power unit cooldown shall be provided.

The following factors shall be taken into account:

- the possibility to compensate any failures of the automatics;

- redundancy of the equipment, pipelines, valves, control points.

Information on the reaction of the system and the RP in case of any emergency situations and failures of the components with due regard for the operator's actions and in the absence of any operator's actions shall be provided.

Information on the reaction of the system, the RP and the entire NPP power unit to any failure without any intervention of the operator in case of the following initiating events shall be provided:

- loss of leak-tightness in the primary circuit entrainment filter;

- failure of the cooling air supply;

- leakage of the coolant;

- blackout.

4. Functioning of the system under external impacts

Any potential emergency modes caused by external impacts shall be specified, in this case the following initiating events shall be considered:

- earthquake - information on the necessity for the system operation during an earthquake with due regard for performance of the prescribed functions by the entire system and its individual components and assemblies, arrangements aimed to ensure the system operation during an earthquake with due regard for isolation of the seismic-resistant part from the non-seismic-resistant one, functioning of the system with due regard for the RP cooldown and refueling or fuel unloading shall be provided;

- aircraft crash - information on the necessity for the system operation in case of the NPP power unit shutdown and cooldown due to the above-mentioned IE shall be provided.


Reliability parameters for the system equipment components from the technical specifications and technical documentation for the equipment shall be presented.

Qualitative analysis and calculation results for the system reliability parameters shall be provided based on the reliability parameters.

Conclusion on the system reliability shall be made on the basis of the calculations and their results.


Compliance with the RD requirements for safety, the design criteria and principles shall be demonstrated.

9.2.2. - 9.2.12. Description and analysis of the systems

In accordance with Section 9 the following systems shall be considered further in subsections 9.2.2 - 9.2.12:

9.2.2. Primary circuit sodium flow control system

9.2.3. System for detection of defective fuel assemblies

9.2.4. Spectrometric control system of the primary circuit sodium

9.2.5. System for cesium content monitoring in the primary circuit sodium

9.2.6. System for cesium removal from the primary circuit sodium

9.2.7. System for oxide content monitoring in the primary circuit sodium

9.2.8. Primary circuit vessels

9.2.9. Primary circuit coolant sampling system

9.2.10. Auxiliary sodium systems of the SFAD

9.2.11. System for transfer of the spent fuel assemblies for storage and transportation from the RP building

9.2.12. System for monitoring of spent fuel assemblies in the shielding chamber

9.3. Other auxiliary systems

Information on the normal operation safety-related auxiliary systems shall be provided:

9.3.1. RP gas cavity blowdown system

9.3.2. Fuel element cladding integrity monitoring system

9.3.3. Gas supply system of the compensator tank

9.3.4. Radiation and dosimetric control system including sampling of radioactive process media

9.3.5. SFP water cooling system

9.3.6. Reliable service water supply system for cooling of the RCP, SFAD and SFP

9.3.7. Plenum ventilation systems

9.3.8. Exhaust ventilation and filtration systems

9.3.9. Air conditioning systems

9.3.10. Ventilation system for the MCR and ECR rooms

9.3.11. Ventilation system for the SNF facility

9.3.12. Ventilation system for the fresh fuel facility

9.3.13. Ventilation systems for the auxiliary facilities and the RW storage rooms

9.3.14. Equipment washing and decontamination system

9.3.15. Communication systems

9.3.16. Lighting systems

9.3.17. Diesel generator fuel storage and supply systems

9.3.18. Diesel generator cooling water system

9.3.19. Diesel generator start-up system

9.3.20. Diesel generator lubrication system

9.3.21. Air intake and exhaust system for the diesel generator unit combustion chamber

9.3.22. Communication systems

It is recommended to follow the description structure given in the section "General requirements" for presentation of information on the auxiliary systems similar to the description of the system for oxide removal from the primary circuit sodium presented in Section 9.2. Specific information shall be presented for each system. The required schemes and drawings shall be attached.

Information in these sections shall not repeat the information contained in any other chapters.

The list of systems in par. 9.3.1 - 9.3.22 may be supplemented, abridged or adjusted in accordance with the particular power unit design.

10. REQUIREMENTS FOR THE SECTION "RADIOACTIVE WASTE MANAGEMENT"

Complete information on handling of radioactive wastes generated in the course of the NPP power unit operation shall be provided. The following shall be considered:

- RW generation sources;

- all possible ways of solid and liquid RW propagation and techniques for their temporary storage;

- all possible ways of gaseous (volatile, aerosol) RW propagation.

The RW management principles shall be stated, and their compliance with the requirements of federal rules and regulations in the area of atomic energy use shall be demonstrated.

10.1. Sources of radioactive waste generation

Sources of RW (radionuclide) generation necessary to develop the RW handling system shall be described, including: sources in the nuclear core (under normal operation conditions and in case of any accidents), the primary coolant circuit, the secondary coolant circuit, the gas circuit of the reactor, the entrainment filters of the primary and gas circuits, the passive emergency heat removal systems, the accident confinement systems, etc.

Results and methodology for calculation of the radionuclide activity in the above-mentioned RW sources shall be presented in a separate subsection, but the basic characteristics of the radionuclides shall be taken from this subsection for calculation of the generated radioactive wastes in order to describe the RW handling process to the fullest extent possible.

The methodology for calculation of the radionuclide activity values in various sources shall be verified, validated, and the global and national experience shall be taken into account.

Analysis of the processes and works (decontamination, repair, etc.) resulting in generation of solid, liquid or gaseous radioactive wastes shall be performed for development of the RW handling

system; these processes and works shall be optimized with due regard for the experience of similar or equivalent NPP power units taking into account the potential RW propagation paths.

Basic information on the amount of radioactive wastes generated in the course of different NPP power unit decommissioning options shall be presented. Quantitative and qualitative RW characateristics shall be substantiated by calculations.

10.2. Gaseous radioactive waste management systems

All systems of the NPP power unit capable of becoming potential sources of radioactive substance releases into the rooms and the environment in the form of GARW, including ventilation systems in the controlled access area of the buildings and the process blow-off treatment system, shall be presented in the form of a block diagram. Description of all systems shall be provided. Options of gaseous waste handling in all operation modes including accidents in the systems under consideration and design basis accidents at the NPP shall be described.


Description of the systems shall be arranged in accordance with the structure presented in the section "General requirements".

The basic safety principles and criteria implemented in the design in the course of gaseous RW handling shall be stated.

Class, category, fire hazard and seismic resistance group shall be specified for the main components of the systems in accordance with the classification given in the effective regulatory documents. Classification data for the system and its components shall be presented in this section from the system information integrity considerations. Reference to any other sections of the report may be given.

It is recommended to present the information in tabular format.

Methodologies and criteria for calculations of the systems shall be presented with expected annual releases of gaseous radioactive substances and expected exposure doses for the workers and the public.

All applied calculation methods and assumptions with due regard for the meteorological and hydrological conditions specified in Section 2 shall be provided.

The adopted design values of radionuclide activity in all units of the systems shall be presented together with the input data for determination of these values and layout of the system equipment for calculations of the biological protection.

It shall be demonstrated that the implemented principles and the corresponding technologies enhance the RW processing efficiency, and the adopted systems contain all state-of-the-art technological achievements intended to reduce exposure doses for the workers and the public.

It shall be demonstrated that the systems have sufficient capacity, efficiency and the necessary redundancy to ensure the required RGAW purification degree and non-exceedance of the permissible RW release limits in all operation modes with acceptable leakage of fuel elements corresponding to the safe operation limit as well as in case of design basis accidents.

Peculiarities of the design including the means aimed to reduce the scope of maintenance, equipment down time, the possibility for RGAW ingress into the rooms, as well as the means aimed to enhance efficiency of the medium purification methods shall be described.

The methods and means provided in the design to control RW releases caused by any potential errors of the operators (workers) and failures of the equipment components into the rooms not belonging to the RGAW handling system and into the environment shall be described. Efficiency of the arrangements for dosimetric monitoring and control of the systems for automatic release limitation in case the release values exceed the established limits shall be substantiated.

All equipment of the systems where explosive concentrations of gases can occur shall be listed, the design pressure values shall be specified, and safety analysis shall be provided for the equipment adopted in the design. The process instrumentation (including gas analyzers), the explosion prevention measures provided in the design and the arrangements aimed to prevent complete loss of leak-tightness due to an explosion shall be described.

Description of each RGAW handling system and gas flow diagrams demonstrating the process equipment, gas movement paths in the system, capacity and efficiency of the system and the corresponding equipment, the redundant equipment and the procedure for its activation shall be provided. For complex multi-functional systems the subsystems divided into independent parts shall be specified with the relevant equipment description. Maximum and normal input gas flow rates and RGAW concentrations for all operation modes and design basis accidents shall be specified for each system in the tabular format or on the diagrams.

Interfaces of the systems and their boundaries with regard to the equipment of different classification groups shall be indicated on the process flow diagrams.

Intrumentation and controls of the system shall be specified.

All existing bypass lines as well as the conditions affecting their usage and the expected frequency of bypass line usage due to downtime of the equipment shall be specified.

The normal operation mode and all other operation modes shall be described. Ventilation systems of each building where appearance of RGAW may be expected shall be described. The following shall be specified in the description: volume of the buildings, expected flow rates in the building ventilation systems and characteristics of the filters. The normal operation mode for each ventilation system and operation peculiarities for different operation modes including design basis accidents shall be described.

Tables with estimated RGAW concentrations in the NPP power unit rooms for all operation modes including design basis accidents shall be provided.

10.3. Liquid radioactive waste management systems

The basic characteristics of LRW handling systems in all operation modes including accidents shall be described in this subsection.

10.3.1. Sources of generation, design basis

The LRW amount shall be determined with reference to the data on generation of radionuclides presented in Subsection 10.1. The places, process-related and routine works, scenarios and processes accompanied with LRW supply to the power unit rooms and releases of radioactive substances into the environment under normal operation conditions and in case of design basis accidents shall be specified.



The purpose and the basic safety principles and criteria implemented in the process flow diagrams shall be specified.

The category, class and group shall be specified for the systems and their main components in accordance with the classification provided in the regulatory documents. Classification data for the system and its components shall be presented from the system information integrity considerations. Reference to any other sections of the report containing the necessary information may be given.

Efficiency of the applied system calculation principles and criteria shall be confirmed by the data with indication of average expected amounts of generated LRW per a year and per the entire NPP power unit operation period and expected exposure doses for the personnel due to their impacts.

These assessments shall include the data demonstrating that the implemented principles and the corresponding technologies enhance the LRW processing efficiency. Compliance with the RD requirements shall be demonstrated for the LRW solidification techniques with regard to their targeting to reduction of exposure for the workers and the public and the solidified LRW quality.

All applied calculation methods shall be specified. The way to consider the site peculiarities (hydrological conditions) specified in Section 2 shall be demonstrated.

Data demonstrating that the developed systems have sufficient capacity, efficiency and the necessary redundancy to ensure the required degree of any discharge purification from radioactive substances in all operation modes and in case of design basis accidents shall be presented.

Peculiarities of the design including the means aimed to reduce the scope of maintenance, equipment down time, the possibility for LRW ingress into the rooms, as well as the means aimed to enhance efficiency of the waste processing methods shall be described. The adopted design values of radionuclide activity in all units of the system shall be specified together with the input data for determination of these values. Layout and geometry of the system equipment shall be presented in order to perform the biological protection calculations.

Potential human errors and failures as well as operational occurrences that can result in unintended and uncontrollable LRW discharges to the rooms and discharges of radioactive substances into the environment shall be analyzed, and efficiency of the developed arrangements and controls, both process-related and through the use of protections, interlocks, instrumentation, etc. shall be demonstrated. The arrangements and controls provided in the design in order to prevent unintended and uncontrollable discharges of radioactive substances into the environment shall be described.

Description of each system shall include process flow diagrams showing the equipment, the design direction of LRW flows, capacity of the system and the relevant equipment components, the redundant equipment. For complex multi-functional systems the subsystems divided into independent parts shall be specified with the relevant equipment description.

Maximum and normal input liquid flow rates and the LRW specific volumetric activity values for all operation modes including design basis accidents shall be specified for each system in the tabular format or on the diagrams. The input data used to determine the above-mentioned values shall be provided.

Separation of the LRW flows, principles of their separation in accordance with physical and chemical properties, radioactivity, etc. shall be described. All possible bypass lines as well as the conditions affecting their usage and the expected frequency of bypass line usage due to downtime of the equipment shall be specified.

Interfaces of the systems and their boundaries with regard to the equipment of different classification groups shall be indicated on the process flow diagrams. Components and assemblies of the equipment and pipelines contaning increased concentrations of radioactive substances shall be

indicated on the diagrams in order to provide the information required for assessments in Section 11.

Normal operation modes and differences in the NPP power unit operation modes including design basis accidents shall be described for each system.

Parameters, assumptions and input data used to calculate the amount of generated LRW with due regard for the part of treated water which may be included into the closed cycle for the purpose of reuse shall be provided.

The expected values of radioactive substance discharges per a power unit and the entire NPP in all operation modes including design basis accidents shall be presented. The discharge values for each sub-system with indication of their concentrations shall be summarized in the table. All RSb discharge points and the discharge dilution coefficients applied in assessment of the LRW specific volumetric concentrations shall be presented.

Parameters, assumptions and input data used to calculate the unbalance water discharges shall be provided. The expected values of unbalance water discharges in all design operation modes including accidents shall be presented, and their maximum specific volumetric concentrations shall be compared with the standard levels for surface water bodies regulated in the RDs.

10.4. Solid radioactive waste management system

The systems for SRW handling in the course of operation including design basis accidents shall be described.

10.4.1. Generation of solid radioactive wastes

The SRW amount shall be determined with reference to the data on generation of radionuclides presented in Subsection 10.1. The places, process-related and routine works, scenarios and processes that result in their generation or can be accompanied with SRW supply to the power unit rooms and to the environment under normal operation conditions and in case of accidents shall be specified.

Block diagrams with the characteristics of works and the SRW supply routes shall be provided.



The basic safety principles and criteria implemented in the process flow diagrams of the systems shall be specified with indication of the particular RD paragraphs on safety.

The category, class, group and type in accordance with the classification provided in the RDs, seismic resistance, the degree of radiation, fire and environmental hazard shall be specified for the systems and their main components. Classification data for the system and its components shall be presented in this section from the system information integrity considerations. Reference to any other sections of the report containing the necessary information may be given.

Criteria and principles of radiation, fire, environmental safety and seismic resistance applied in the design of the SRW handling systems shall be specified, their impact on the block diagrams of the systems, selection of their components, selection of the storage techniques, SRW characteristics (maximum and expected amounts and volumes, composition and activity of wastes, duration of temporary and long-term storage, the possibility for their inclusion into the natural systems), etc. shall be presented.

Description of each system shall include descriptions of the SRW handling sub-systems used for regeneration of filters, conditioning, operability recovery for the traps, etc. The main parameters of the systems shall be specified and substantiated: capacity, efficiency, fire and explosion safety, degree of protection in case of design basis accidents.

The input data, maximum and expected amounts of SRW, their physical form, composition, source of the wastes (place, process, etc.), radionuclide composition and specific activity shall be presented in tabular format or in any other compact graphical form. Methods to be used for processing (concentration, regeneration, decontamination, storage, disposal, etc.) of each type of wastes, the waste packaging types, the final forms of conditioned wastes and the places of their location shall be described.

The process flow diagrams for processing of each SRW type, the expected configuration of each line and capacity of the equipment, any possible malfunctions and their consequences shall be provided.

Process controls and instrumentation shall be described. Process flow diagrams with indication of interfaces between the systems, the boundaries between the equipment of different classification groups and the instrumentation shall be presented.

Schemes of the packing, storage, loading and transportation areas for different categories of wastes shall be provided.

The arrangements provided in the design in order to prevent ingress of radioactive substances into the NPP power unit rooms and into the environment during normal operation of the SRW handling systems and in case of any accidents shall be described.

Efficiency of the developed arrangements for prevention of SRW ingress to the rooms and the environment as well as the control and monitoring systems shall be demonstrated.

Peculiarities of the design including the means aimed to reduce the scope of maintenance, equipment down time, the possibility for SRW ingress into the rooms, as well as the means aimed to enhance efficiency of the waste processing methods shall be described. The adopted design values of SRW in all units of the systems shall be specified together with the input data for determination of these values. Layout and geometry of the system equipment shall be presented in order to perform the biological protection calculations.

All possible human errors capable to result in unintended and uncontrollable discharges of solid radioactive wastes into the rooms and the environment shall be analyzed, and efficiency of the developed technical and administrative measures shall be demonstrated.

The monitoring systems for the processes, releases and discharges as well as SRW supplies shall be described in par. 10.5.1.

The SRW handling systems intended for treatment of contaminated work clothes, equipment, tools, filters of the ventilation systems as well as any other pressed and non-pressed wastes shall be specified. Maximum and expected input data for the above-mentioned wastes with indication of the waste sources, amounts, radionuclide and chemical composition and activity shall be presented in tabular format. The input data used to determine the applied values shall be provided. The method for conditioning and packing of wastes, the equipment used for this purpose, techniques for processing and packing of large-scale SRW (the reactor nuclear core components, etc.), containers to be used for packaging of wastes shall be described. Compliance of the protective characteristics of the containers with the effective standards and rules shall be demonstrated. Measures provided for decontamination and transportation of containers with wastes to the storage facilities shall be described together with the analysis of any potential operational occurrences including accidents (loss of integrity of the containers with wastes in case of falling, etc.). Arrangements provided for collection of wastes and

decontamination techniques in case of any loss of container integrity shall be described. Safety precautions taken in the course of waste storage prior to loading and transportation and the expected duration of SRW storage at the site shall be presented. Schemes of the packing, storage, loading and transportation areas shall be provided. Maximum possible and expected annual amounts, radionuclide composition and activity shall be specified for each SRW category subject to removal from the site.

10.5. Radiological control

The system ensuring radiological control in the course of RW management shall be described (references to the information presented in Section 11 may be given) including the sampling sub-system in the course of gaseous, liquid and solid RW handling and RSb releases and discharges in all operation modes, emergency situations and accidents.


The basic radiation safety principles and criteria implemented in the design and (or) process flow diagrams of the systems shall be specified with indication of the particular RD paragraphs on safety.

The category, class, group, type, etc. in accordance with the classification provided in the regulatory documents as well as seismic resistance, degree of radiation hazard, etc. shall be specified for the systems (and for their main components in case of necessity). Classification data for the system and its components shall be presented in this section from the system information integrity considerations. Reference to any other sections containing the necessary information may be given.

Objectives, principles and criteria shall be specified, and the way to use them in the design of the entire system and its individual sub-systems shall be demonstrated. Differences in the sub-systems intended for functioning under normal power unit operation conditions, in case of any emergency situations, design basis and beyond design basis accidents shall be specified.

Purpose of the systems shall be specified, their block diagrams shall be provided, and their operation principles shall be described.

The data characterizing the following aspects shall be provided:

- reliability and sufficiency of measurements for all operating conditions of the systems;

- degree of protection against unauthorized access to the stored information;

- sufficient redundancy of the system components under normal operation conditions and in case of their functioning under extreme conditions;

- sufficiency of the primary detector locations;

- correct selection of the sampling points and sufficiency of their number in order to provide correct monitoring of the media composition;

- sufficiency of the emergency situation warning means, their correct locations and substantiation of the alarm setpoint selection.

The following information shall be also provided for radiological monitoring of the waste management processes and "flows":

- location of sensors;

- types of sensors, characteristics, type of measurements;

- instrumentation and control devices, redundancy, independence of performed measurements;

- RSb concentration measurement range and input data for determination of the provided range;

- types and locations of the warning devices, radiation level alarms (particularly emergency ones) and controllers and their description;

- redundant power supply;

- setpoint values for emergency alarms and activation of protections, interlocks and controllers; the input data for determination of these values;

- description of the measures provided for calibration, maintenance, verification, decontamination and replacement of control instruments. The following information shall be presented for each sampling device:

- the basis for selection of sampling point locations;

- expected flow, composition and concentration of radioactive and chemical substances in the samples;

- frequency of sampling, type of the sampling equipment and methods used to obtain representative samples;

- laboratory analysis methods and sensitivity of the instruments.

11. REQUIREMENTS FOR THE SECTION "RADIATION SAFETY"

Criteria of radiation safety assurance for the workers and the public (in accordance with the dose limits, surface contamination of various surfaces, human hands, releases and discharges of radioactive substances) under normal operation conditions and in case of any accidents shall be provided in this section.

It shall be confirmed that the individual dose limits for the workers in all normal operation modes and in case of design basis accidents will not exceed the established values, and ingress of radioactive substances into the environment will not result in exceedance of the basic dose limits for the public established in the regulatory documents.

The programs for the radiation situation monitoring in the rooms, the individual dosimetric monitoring program and the environmental radiological control program shall be provided.


- methods for protection against external exposure (gamma rays and neutrons from the nuclear core, structural materials of the reactor, reloaded fuel assemblies and equipment containing radionuclides);

- methods for protection against internal exposure (peroral and inhalation intake).

The degree of compliance with the requirements of the effective regulatory documents for radiation safety shall be specified for each subsection.

Special references to the information presented in other sections may be given (in case of necessity).

11.1. Radiation safety concept

Principles, criteria, calculation methods, engineering features and administrative measures used to ensure protection of the workers, the public and the environment against impermissible impact of radiation and toxic compounds (due to radioactive sodium) shall be described.

It shall be demonstrated that compliance with the requirements for safety assurance is justified by the operating experience for similar NPP power units and will not result in any exceedance of the exposure levels regulated in the RDs. Impact of hazardous factors in all power unit operation modes and in case of accidents shall be limited to the lowest reasonably achievable levels with due regard for the economic and social factors. Achievable exposure levels shall be presented in the form of the collective annual dose (quota) for the workers and the public and the annual dose for individual categories of the workers under normal operation conditions and in case of design basis accidents.

The radiation protection principles, selection of the technical and organizational solutions used in the design of the power unit reactor components in order to reduce the radiation exposure level, particularly chemical exposure due to radioactive sodium, to the lowest reasonably achievable level with due regard for the economic and social factors (ALARA principle) shall be described.

Technical and organizational solutions aimed to reduce the exposure level for the workers shall be described:

- arrangement of the biological protection shields;

- arrangement of closed circuits with radioactive media;

- arrangement of controlled discharge and treatment of potential radioactive leaks, etc.

The plant zoning criteria, the design solutions for protection of the workers in the course of operation and in case of beyond design basis accidents, for example by limitation of external and internal exposure shall be described.

The way to use the accumulated experience in design and operation of other power units in order to reduce the exposure levels to the lowest possible values shall be demonstrated.

Any means provided in the design in order to reduce exposure levels in the controlled access area rooms and to reduce the time of the personnel stay in these rooms, particularly to decrease the number of RSb sources, to enhance the radiation protection, to reduce the time of maintenance, to facilitate access to the equipment, to simplify operational procedures and also to reduce and simplify any other actions required within the operation period shall be described.

The list and brief characteristics of the rooms belonging to the controlled access area as well as the list of special technical solutions aimed to ensure compliance with the RSS requirements shall be provided.

11.2. Design basis

Radiation criteria used in the development of manuals and engineering features for performance of radiation-hazardous works, including maintenance, in-service inspections, metal state control, refueling of the reactor nuclear core, works with radioactive wastes in order to reduce the exposure doses in accordance with ALARA principle shall be specified.

The ways to limit internal and external exposure for the workers and to arrange separation of workplaces and rooms in accordance with the zoning principle shall be presented.

The list and quantitative values of the radiation parameters such as total specific activity of fission products in the primary circuit coolant, specific volumetric activity of air in periodically attended rooms, contamination levels for the surfaces of the rooms and equipment installed in the periodically attended rooms, etc. shall be provided.

The list and quantitative values of the process criteria to be complied with in order to maintain the exposure for the workers at the lowest reasonably achievable level (for example, the operation limits for damage of fuel elements, the coolant leakage values, etc.) shall be provided.

11.3. Radiation sources

11.3.1. Equipment contaning radioactive substances

Information on the content of radioactive substances in the equipment components (except for the RW handling systems described in Section 10) representing sources of radiation taken into account in the biological protection calculations and design shall be provided. The following shall be described:

- the reactor nuclear core as the source determining the ionizing radiation levels in the course of the reactor power operation in the rooms behind the biological protection where any workers can be present as well as the source of fission products supplied to the primary circuit;

- materials of the reflector and other structural components of the reactor as the source of capturing and activation gamma radiation;

- the primary circuit as the source of activation products for the primary circuit coolant and activated corrosion products as well as fission products entering the coolant due to any defects of fuel element claddings;

- the secondary circuit and other systems and equipment of the power unit that can contain radioactive substances;

- the system of refueling, SNF storage and transportation containing fission products in the irradiated fuel and structural material activation products;

- other radiation sources including start-up neutron sources, sources for calibration of instruments and devices, sources for gamma-radiography, nuclear reaction by-products and any other sources requiring radiation protection.

Description of the radiation sources (except for the reactor nuclear core) shall contain the table of radionuclide composition, information on the activity, geometric parameters of the source as well as the input data for determination of the specified values.

It should be substantiated that any ingress of fission products into the coolant in the course of power operation does not exceed the permissible operation limit for damage of fuel elements. Increase of the fission product ingress to the coolant from the fuel in case of any emergency situations and transient modes shall be taken into account.

The information shall be presented in such a way so that to serve as the input material for calculations of the biological protection.

11.3.2. Sources of gaseous radioactive substances

The sources of gaseous RSb ingress into the atmosphere of the controlled access area rooms taken into account in development of protective measures and assessment of occupational exposure doses

shall be described. Apart from the sources existing under normal operation conditions, the information on the sources appearing due to failures of the main equipment and in the course of repair works (opening of the reactor, SNF transportation, etc.) shall be provided.

Description shall include the calculation results with regard to concentration of radioactive gases and aerosols expected in the course of normal operational and transient modes, in case of any operational occurrences and accidents.

Models, parameters and input data required to calculate concentration of radioactive gases and aerosols shall be provided.

11.4. Design peculiarities with regard to radiation protection

11.4.1. Location plan and layout of the buildings, structures and equipment

The plan (scale 1:1000) of the complex of industrial buildings, facilities and rooms of the NPP power unit with layout of the process equipment representing the source of radiation as well as all radiation sources specified in Subsection 11.3 and Section 10 shall be provided. The planning and layout concept for the buildings, facilities and equipment of the buildings and facilities from the viewpoint of radiation protection shall be presented.

The following shall be indicated on the plan:

- boundaries of the controlled access area and classification of its rooms into non-attended, periodically attended and attended ones, as well as rooms of the uncontrolled access area including the administrative and amenity building;

- location of the personnel airlocks, stationary decontamination locks, active laundries, medical posts;

- schemes of the personnel and transport movement, delivery of clean equipment and materials and removal of contaminated ones;

- location of the places for storage of contaminated equipment, decontamination areas, places for solid RW collection, control panels for the equipment and mechanisms of the RW processing systems;

- location of sensors and control panels of the radiological control system;

- location of laboratories for analysis of the radioactive media samples (chemical, radiochemical, radiometric, spectrometric), the laboratory for personal monitoring as well as the laboratory for monitoring of metals, the repair and calibration laboratory (workshop), storage facilities for ionizing radiation sources;

- location of the external dosimetry laboratory, observation stations and checkpoints;

- places for collection of non-radioactive wastes and arrangement of control in order to eliminate accidental ingress of radioactive substances into non-radioactive wastes.

Classification of the NPP power unit zones and rooms adopted in the design and used as the basis for design of the biological protection against penetrating radiation and prevention of radioactive contamination of the air in the attended rooms of the controlled access area shall be presented.

11.4.2. Structural peculiarities of the systems and equipment components

Design peculiarities of the equipment and plants enabling to reduce the occupational exposure

doses in accordance with ALARA principle shall be specified.

The description shall include the structural peculiarities reducing maintenance or other operations in radiation fields, decreasing intensity of the sources and also providing quick entrance to the room of the building or facility, easy access to the workplace, remote performance of the operations, reduction of the personnel stay time or any other measures aimed to reduce exposure of the workers.

Description of the methods used in the design in order to reduce generation, distribution and accumulation of the activated corrosion products, application of the materials with the minimum cobalt content in the primary circuit, compliance with the optimal coolant chemistry regimes, minimizing of the dead areas (cavities, pockets) where activation products can accumulate shall be included. Illustrative examples shall be provided including equipment drawings and piping schemes for the components requiring access of the personnel during the power operation of the unit (equipment of the active water treatment systems, tanks, coolers, pumps, SG, sampling systems). Location of the sampling points, instrumentation, control panels and stations shall be indicated.

11.4.3. Biological protection

Information on the biological protection shall be provided for each radiation source described in Section 10 and Subsection 11.3, including characteristics of the protective materials, thickness of coatings, methods for determination of the protection parameters, geometrical parameters of the source and protection.

Special-purpose protective devices and equipment including containers, casings, screens, loading equipment, etc. used for handling of any RW type shall be specified.

Calculation programs used for calculations of the protection shall be provided. Results of the calculations shall be presented including the design radiation level in attended and periodically attended rooms of the controlled access area as well as in the rooms of the uncontrolled access area, particularly the administrative and amenity building, in the course of normal operation, in case of any design basis accidents and during the NPP power unit decommissioning.

11.4.4. Ventilation, filtration and conditioning systems

The main design parameters for the ventilation systems of the controlled access area shall be described, including the repair ventilation, as well as any components intended to ensure safety of the workers and belonging to the ventilation systems, but not described in Sections 9 and 10. Removal of gas and aerosol fission products from the rooms of the controlled access area, the process blow-off as well as the RW release monitoring system shall be described in Section 10.

The principle of separate ventilation for the rooms of the controlled and uncontrolled access area shall be described.

Examples shall be presented to illustrate the measures provided in the design for air purification from radioactive gases and aerosols including the plan of the rooms where purification is performed and purification devices (filtering stations) are installed, the piping schemes and filter valves.

The maintenance conditions for the ventilation, filtration and conditioning system shall be specified, and the means for monitoring, testing and isolation of the systems shall be also described. The means for determination of air purification efficiency, replacement and transportation of used filtering elements shall be described. Characteristics of the applied air purification means and criteria establsihed for replacement of the filtering elements shall be specified. Purification coefficients adopted for the radiation safety analysis shall be specified. Due to dependence of these coefficients from the filtration conditions they shall be assumed for the radiation situation assessment based on the most severe

operation conditions for the filtering systems (design sizes of the aerosol particles shall be assumed to be equal to the size of the most penetrating particles for each filter; the most unfavorable temperature and humidity conditions out of all possible variants shall be assumed for iodine filters and gas sorbents).

11.4.5. Radiological dosimetric control system

1. Radiological dosimetric control system Sampling of radioactive process media

Criteria for selection of the engineering features for radiological control, development of the scheme of sampling points for radioactive process media and the external environment samples and location of the equipment shall be provided. The engineering features provided in the design for radiological control shall be described, including the following hardware:

- continuous monitoring devices based on stationary automated systems and stationary instruments;

- in-process control devices based on portable and mobile instruments and units;

- laboratory analysis devices based on laboratory units, means for sampling and preparation of radioactive samples for analyses;

- individual monitoring of the workers.

The list of radiological control objects, classification of the control types in accordance with the requirements of GSR and NPP SR shall be provided, and the seismic category and fire resistance category of the system and its equipment components as well as the system category according to its purpose shall be specified.

Description shall include the main technical characteristics (controlled parameters, types and number of sensors, the measurement range, the basic error), information on the metrological support methods and means, information on the alarm units, recording devices and location of sensors, indicating (reading) and signaling devices. The schemes of sampling lines with valves shall be presented.

Design options of the hardware, the seismic impact it is designed for, compliance with the requirements for fire, electrical and mechanical safety shall be specified.

Location of the air sampling points for control of gas and aerosol activity shall be indicated, the air sampling system shall be described, criteria and methods used to obtain representative results of the radioactive gas and aerosol concentration measurement shall be provided.

The capabilities of the radiological control hardware to measure the radiation situation parameters, particularly intensive radiation and exposure doses for the workers in case of a radiation accident shall be described, and the necessity for instrumentation in order to perform these measurements shall be substantiated.

The list of equipment aimed to monitor contamination of skin, clothes, equipment and various surfaces with alpha-active substances shall be provided; the issues related to monitoring of the radiation parameters in the course of fresh NF loading and SNF unloading shall be also reflected, and the list of controlled radiation parameters shall be presented.

Software tools for processing and presentation of information, the programs ensuring prediction of the radiological consequences of any events at the NPP, collection, storage and systematization of data on radiation pollution of the environment and exposure doses for the workers and the public shall be described.

2. Environmental dosimetric control system

The purpose and configuration of the environmental ARSMS shall be presented.

Configuration and equipment of the stationary external dosimetric laboratory and the mobile laboratory shall be described.

Location and equipment of the stationary observation stations and the environmental radiation situation measurement points within the sanitary-protective area and the supervised area shall be specified.

11.5. Assessment of dose commitments under normal operation conditions and in case of accidents

Annual duration of the personnel stay in the rooms of the controlled access area under normal operation conditions, in transient modes and in the course of repair works shall be assessed (including the number of people).

Duration of the personnel stay (in man-hours) and the RSb ingress to the human organism through inhalation shall be assessed for the rooms of the controlled access area described in par. 11.3.2 where gas and aerosol activity is expected.

The annual individual dose (total and separate for external and internal exposure) and dose commitments of the workers (collective dose) in the course of such basic functions as operation, maintenance, in-service inspection and examination of weld joints, RW handling, refueling of the reactor nuclear core, repair works shall be assessed.

The input data, calculation methods and models and assumptions applied to determine the above-mentioned values shall be specified. In case the estimated (predicted) exposure doses and dose commitments are unacceptably high the arrangements provided in the design for their reduction to acceptable values shall be described.

Information on the exposure doses for the workers obtained during operation of similar NPPs may be used to assess doses and dose commitments in the course of any unpredictable operations with due regard for certain conservative assumptions.

The potential annual dose value at the boundaries of the controlled access area, the uncontrolled access area (the industrial site) and the NPP SPA as well as in the location areas of the main radioactivity sources at the NPP territory (power units, RW storage facilities, points of radioactive releases and discharges, etc.) shall be assessed. The annual exposure dose for the construction workers from these sources at the operating NPPs in the course of construction shall be assessed. The input data, calculation methods and models and the adopted assumptions shall be specified,

The exposure dose for the workers in case of design basis accidents (and the exposure dose for construction workers) shall be assessed. The input data, calculation methods and models and the adopted assumptions shall be specified,

11.6. Radiological control arrangement and programs

11.6.1. Arrangement

The organizational structure of the operating organization departments including the NPP radiation monitoring service ensuring implementation of the program shall be presented.

Organizational and administrative measures for control of the personnel stay in the controlled access area and compliance with the manuals for performance of radiation-hazardous works shall be described.

Information on any mobile services equipped with the hardware aimed to obtain the radiation situation data both under normal operation conditions and in case of any emergency situations and design basis accidents shall be provided.

The organizational structure of the system and storage conditions for the radiological control instruments, their calibration and metrological validation shall be described.

The way to inform the state safety regulating authorities on the results of the program implementation shall be demonstrated. Reference to the information presented in Section 13 may be given in this section.

11.6.2. Radiological control programs

The radiological control programs for normal operation and accidents shall be provided. The list of issues to be considered in each section of the control programs is given in Appendix 4.

Radiological control program for the power unit

The radiological control program for the power unit shall include the following sub-programs: integrity monitoring for the barriers in the way of radioactive substance and ionizing radiation propagation, control of exposure for the workers; RW management control, control for non-propagation of radioactive contaminations.

1. Sub-program of integrity monitoring for the barriers in the way of radioactive substance and ionizing radiation propagation

The sub-program shall contain the information sufficient to:

- assess integrity of the barriers;

- assess reaching of the regulated intervention levels (operation limits and safe operation limits for safety barriers);

- ensure independent and prompt informing of the state safety regulating authorities on the integrity and state of the barriers.

2. Sub-program of the exposure control for the workers

The sub-program shall include description and substantiation of the radiation situation monitoring scope at the places of potential personnel stay and individual dosimetric monitoring sufficient for:

- determination on exposure dose rates in attended, periodically attended and non-attended rooms (for the latter - within the period of repair with the power unit shut down);

- determination and assessment of equivalent exposure doses for the workers within the entire range of potential radiation exposure levels occurring under normal operation conditions as well as in case of design basis and beyond design basis accidents (the same shall be determined for the public in case of beyond design basis accidents);

- obtaining of information for emergency assessment of the radiation situation in the areas of the personnel stay for timely selection and implementation of optimal protection measures during development of design basis and beyond design basis accidents.

3. Sub-program of the radioactive waste management control

The sub-program shall contain description and substantiation of the radiological control scope for handling of liquid, solid and gaseous wastes as well as releases and discharges. In this case it should be demonstrated that the control scope is sufficient to:

- obtain the information on the radiation situation created by radioactive releases and discharges to the environment, and determine the exposure doses for the personnel at the power unit, in the sanitary-protective area and the public in the supervised area;

- determine the amounts and radionuclide composition of the radioactive wastes generated and stored at the NPP power unit;

- obtain the information on the dose burden for the personnel occurring in the course of RW handling works;

- detect and record any exceedance of the established values for radioactive releases and discharges to the environment, as well as unauthorized RW movement and accumulation at the NPP site.

4. Sub-program of control for non-propagation of radioactive contaminations

The sub-program shall contain description and substantiation of the radiological control scope for efficiency of the barriers preventing propagation of radioactive substances to the environment sufficient to:

- determine the RSb contamination levels for surfaces of the process rooms and equipment, skin, footwear, working clothes, personal protection equipment of the workers and any vehicles upon crossing of the controlled access area boundaries;

- determine the RSb contamination levels for personal clothes and footwear of the personnel upon crossing of the NPP territory boundaries;

- determine the RSb contamination levels for vehicles and transported cargoes upon crossing of the NPP territory boundaries.

Program for radiological monitoring of the environment in the sanitary-protective area and the supervised area

The program shall contain description and substantiation of the radiological control scope in the NPP SPA and supervised area with regard to radioactive contaminations of the environmental media and exposure for the workers and the public sufficient to:

- obtain the information for assessment of exposure for the critical groups of the public and the personnel;

- obtain the information for assessment of tendencies and changes of RSb accumulation in the environmental media and human organism;

- obtain the information for emergency assessment of the radiation situation in any territory subjected to radioactive contamination during a beyond design basis accident in order to establish the radiation accident zone boundaries and to implement the required measures for protection of people and the environment (character of intervention) with due regard for the fact that the proposed intervention shall do the society more good than harm.

Radiological control programs in case of emergency situations and accidents

The program shall contain description and substantiation of the radiological control scope at the power unit in case of any emergency situations, design basis and beyond design basis accidents (with due regard for potential accident development scenarios including ignition of the sodium coolant) as well as control of any radiological accident through the use of the NPP efforts and resources in coordination with the radiological monitoring performed by the first-priority UNARSMS institutions and stations at the territory of Russia sufficient to:

- detect any loss of integrity in the barriers;

- detect the amount and radionuclide composition of the released (discharged) radioactive substances;

- ensure sampling of air from the rooms after the accident onset;

- determine, assess and predict the radiation situation in the rooms, at the NPP site, in the sanitary-protective area and the supervised area;

- determine, assess and predict the equivalent external and internal exposure doses for the workers and all persons within the site boundaries and in the SPA, as well as for the critical groups of the public within the supervised area;

- determine the boundaries for the zone of emergency measures, the zone of preventive measures and the restriction zone within the radiation accident zone;

- predict the possibility for reaching of the intervention levels and establish the emergency preparedness levels;

- guarantee functioning of the radiological control system section under the conditions of a considered beyond design basis accident with the most severe radiation situation;

- develop and implement the optimum measures for protection of the workers and the public;

- predict the radiation situation in the environment along the path of radioactive release propagation in the atmosphere during development of a beyond design basis accident for immediate protection of the public with due regard for the established criteria for implementation of the public protection measures;

- ensure timely informing of the local authorities on the necessity to be ready for implementation of the public protection arrangements.

11.6.3. Medical care and health protection for the personnel

1. Medical care organization

The organizational structure of the medical support and health monitoring for the workers related to prevention and reduction of harmful radiation impacts shall be presented.

2. Equipment, protective means and devices

Location of the medical and sanitary facilities (health centers, medical aid posts, active laundries) and the types of equipment (instruments, hardware) for sanitary control shall be specified. Personal protection equipment, its characteristics, usage and maintenance shall be described.

Location of the main equipment ensuring radiation safety of the workers (including changing rooms, shower rooms, the rooms of duty dosimetricians and the outgoing dosimetry control posts), laboratory units for radiometric and spectrometric analysis, storage facilities for protective clothes,

respiratory protection devices, decontamination equipment shall be specified.

3. Methods of radiation protection

Methods of special-purpose air sampling as well as selection and use of special-purpose equipment and devices for respiratory protection shall be described.

Criteria and methods of radioactive contamination control for the personnel, equipment and surfaces shall be described.

12. REQUIREMENTS FOR THE SECTION "SAFETY SYSTEMS"

Substantiation of the SS selection, their functional purpose, classification of the SS systems and components, principal schemes of the safety systems and the basic structural peculiarities of SS components as well as substantiation for performance of the safety functions assigned to them shall be provided in this section. In case any individual safety systems are presented in other sections the name of the system shall be specified in the text, and the reference to the section with full description shall be given.

The SS description presented in any other section may be completely repeated in Section 12.

12.1. List of safety systems

The list of all safety systems provided at the NPP power unit shall be presented.

The list shall include the following basic safety systems:

- the reactor shutdown systems;

- the systems for emergency heat removal from the reactor;

- the overpressure protection systems (for the primary, secondary and third circuit);

- the SG emergency protection system;

- the systems for protection against coolant loss in case of any loss of leak-tightness in the pipelines and equipment of the primary circuit auxiliary systems;

- the overpressure protection systems for the SFAD casing;

- the safety containment of the reactor;

- the safety containments for the pipelines of the primary circuit auxiliary systems;

- the safety containments for the pressure pipelines of the primary circuit and the safety shell of the pressure chamber;

- the safety containment for the spent fuel assembly drum;

- the safety containments for the pipeline sections from the SFAD to the overflow vessel;

- the melted fuel catcher;

- the ventilation system for the rooms with the sodium systems of the primary circuit and the SFAD (fire ventilation);

- the reactor ptotective shroud;

- the sodium burning suppression systems;

- the fire extinguishing system for the rooms with sodium systems;

- the sodium aerosol filtration system;

- the SFP lining (external);

- leak-tight rooms;

- the reliable water supply system for the PSSs;

- the emergency power supply system;

- the ventilation systems for the CSS rooms;

- the fire extinguishing systems for the SS cable rooms;

- the ERCS AHE air supply system;

- the SS fire extinguishing systems;

- the sodium leakage detection systems;

- the sodium burning suppression systems (including envelope structures of the rooms);

- the CSS for actuation of safety systems;

- the CSS;

- the information support system for the operator.

12.2. Protective safety systems

The list of protective safety systems shall be provided, and each of them shall be described.

The information in description of the systems shall be presented in compliance with the following structure.

12.2.1. Purpose

Information on the purpose of each safety system and their components shall be provided with indication of the performed functions and safety class in accordance with the GSR requirements, the safety group in accordance with the NPU Rules, the seismic category in accordance with the Design standards for seismic-resistant nuclear power stations with substantiation of this classification.

Other regulatory documents applicable to the system or component shall be also listed (in case of necessity classification in accordance with these documents shall be specified).

12.2.2. Design basis

Information on the design basis, requirements and design criteria shall be provided.

12.2.3. Description of the design and (or) the process flow diagram

The following information on the systems and their components shall be provided:

- description of the design and (or) the process flow diagram

- detailed (but without excessive details) figures (schemes) illustrating the system design or its process flow diagram. All components listed in the description of the design and (or) the process flow diagram shall be shown as individual positions on the figures (schemes);

- external conditions and ambient parameters affecting the PSS components in all operation modes;

- basic technical characteristics of the system.

Protection of the systems against external impacts (fires, falling of any objects, flooding, etc.) shall be described.

The way to protect the systems against unauthorized intervention of the workers shall be demonstrated.

In case the systems include any pipelines, valves, heat exchangers, pump sets, tanks, safety valves and other equipment the basic information reflecting peculiarities of these components shall be provided in their descriptions.

12.2.4. Materials

Information confirming compliance of the materials, manufacturing and control methods with the requirements of the NPU Rules, the Strength Calculation Standards, the regulatory document "Equipment and pipelines for nuclear power units. Welding and surfacing. Basic provisions", the regulatory document "Equipment and pipelines for nuclear power units. Welded joints and overlaying. Control rules" shall be provided.

In case the selected material is not specified in the NPP Rules or is specified but used with certain deviations from the application conditions stated in these Rules the reference to the documents substantiating the possibility for application of the selected material shall be given.

12.2.5. Design substantiation

It shall be demonstrated that all system components are designed with due regard for the possibility to withstand the ambient conditions (pressure, temperature, vibration, shock impacts, humidity and radiation fields occurring in the course of operation, etc.) both during normal operation and in case of any operational occurrences including accidents.

Information on the calculations performed in order to substantiate the SS design and information on the SS compliance with the safety assurance requirements shall be provided (with reference to Section 15).

Information on the protective safety systems and consideration of the similar PSS operation experience in the system design shall be presented.

Information on the S&RW and R&DW performed to substantiate the design shall be presented in accordance with the following scheme:

- the list of performed experimental works;


- results of the experiments with conclusions.

12.2.6. Quality assurance

The way to assure quality of all system components in the course of manufacturing, installation and construction shall be demonstrated.

12.2.7. Control

The signals used to initiate the system, the required sources of energy and working medium shall be listed.


- the list of measurement points;

- the list of protections and interlocks (in-system);

- operation algorithms, alarms;

- description of the monitoring systems, accuracy of the parameter determination;

- the list of manual operations for control of the systems;

- availability of the operator support means in the control of systems and components.

12.2.8. Control and testing in the course of operation

The following information on the systems at the operation stage shall be provided:

- frequency of the state control and testing of the systems and components;

- regular inspections of the systems and their components.

Information on the metal state control for the system pipelines and equipment shall be specified.

12.2.9. Commissioning

Information on the commissioning works for the system including its testing shall be provided. Sufficiency of the pre-operational testing for the NPP power unit safe operation assurance shall be substantiated.

12.2.10. Functioning of the system

The following shall be described: functioning of the system including transient modes in case of scheduled start-ups and shutdowns, state of the system and its components, their interactions among themselves and with other systems during performance of the prescribed functions.

The list of signals requiring activation of the particular system shall be given. The functions of the

system shall be determined for each mode, the criteria for performance of the functions assigned to the system shall be specified, and the impact of the mode on the state of the system and its components shall be demonstrated.

12.2.11. Functioning of the system in case of failures

Failures of the system components including human errors shall be analyzed, the impact of failures on the system operability and the possibility to perform the prescribed functions shall be assessed. In this case failures of the passive and active components, instrumentation of the system itself and the associated CSSs and SSSs shall be described. Special attention shall be paid to the analysis of common cause failures.

12.2.12. Reliability

Information on the system reliability analysis and calculation shall be provided.

12.2.13. Design assessment

Compliance with the design basis specified in par. 12.2.1 shall be demonstrated.

Compliance of the PSS design and its implementation with the RD requirements for safety shall be acknowledged.

12.3. Localizing safety systems

Each localizing safety system shall be considered.

The information in description of the systems shall be presented in accordance with the same structure and scope as recommended for the PSS description. Besides, any additional information with due regard for peculiarities of the LSSs shall be provided, namely:

- the time from onset of the design basis accident the LSS is aimed to protect against to the moment when access of the workers to the accident confinement zone becomes possible shall be specified in par. 12.2.1;

- information on the measures provided in the design in order to bring the systems into the initial state after performance of the prescribed functions shall be presented;

- information on the stability of the applied materials and their coatings under accident conditions shall be provided in par. 12.2.11.

12.4. Supporting safety systems

Each supporting safety system shall be considered.

The information in description of the systems shall be presented in accordance with the same structure and scope as recommended for the PSS description. Besides, any additional information with due regard for peculiarities of the SSSs shall be provided, namely:

- information on the operation duration (limited or unlimited) in the emergency period shall be presented in par. 12.2.3. Information of the required stock of expendables shall be provided. Information on the system filling and makeup (volumes, flow rates in the course of filling and makeup)

shall be provided;

- information on the supporting safety systems and characteristics of the places that may be used to activate the system and its individual components shall be provided in par. 12.2.7;

- information on the stability of the applied materials and their coatings under normal operation and accident conditions shall be provided in par. 12.2.11. Special attention shall be paid to generation of any secondary decomposition products posing hazard from the viewpoint of toxicity and explosiveness under any system conditions different from the design ones. For example, decomposition process for diesel fuel, freon, electric insulation of the cables, etc. in case of a fire shall be considered.

The EPSS SSS is based on two technologies: electrical and thermomechanical, and consists of two almost independent parts (the electrical part which is called the EPSS in accordance with the established practice, and the thermomechanical part - the SDGS). Description of the electrical part shall be provided in Subsection 8.4, Section 8.

The SDGS description shall be presented in accordance with the structure recommended in Subsection 8.4.

Information on the priority of protection commands for the diesel electric sets and initiating commands for the CSS shall be also specified in Subsection 12.2.7.

12.5. Control safety systems

12.5.1. CSS for actuation of safety systems

The CSSs intended for actuation of safety systems (protective, localizing and supporting) and the own CSSs of safety systems shall be considered. Description of the step-by-step start-up programs shall be provided where only their peculiarities and differences from the CSs of the normal operation systems shall be considered.

Complete information shall be presented in description of the CSSs intended for SS initiation in accordance with the same structure and scope as recommended for the PSS description.

Block diagrams of the CSS command generation shall be provided. The list of all sensors (except for NFME) with indication of the systems where they are installed shall be provided.

It should be noted that all sensors are installed in safety-related normal operation systems but shall comply with the RD requirements for SS safety.

Compliance of the CSS with the GSR and NPP RF NSR requirements shall be reflected.

12.5.2. CSS functioning

Peculiarities of the CSS functioning shall be provided.

12.5.3. Information support of the operator

Information on the operator support system in monitoring of the power unit safety conditions and data on the information support of the operator in case of beyond design basis accidents shall be presented. It shall be demonstrated that control over the progress of the prescribed safety function performance is maintained in case of any beyond design basis accident (considered in Section 15).

13. REQUIREMENTS FOR THE SECTION "COMMISSIONING"

Information on arrangement, scope, sequence and time limits of the commissioning works and testing performed in the course of the NPP power unit commissioning for all safety-related facilities, equipment, systems and components of the NPP power unit shall be provided.

The information shall cover all stages of the commissioning, beginning from acceptance of the equipment and systems after installation and finishing with the integrated trial of the NPP power unit at the rated power and its delivery for commercial operation (including such works as pre- and (or) post-installation cleaning of the equipment and circuits, functional adjustment and testing of individual equipment units and valves, as well as the entire systems; integrated testing of the RP equipment, initial loading of NF to the nuclear core, reaching of the first criticality and the specified minimum power level; step-by-step power increase up to the rated value and commissioning of the NPP power unit).

It is reasonable to use the experience in preparation of the documentation developed previously for commissioning of BN-350 and BN-600 power units, reporting documents on the commissioning works and testing performed during commissioning of these units as well as standard documentation applicable to other power unit types.

13.1. General provisions

The basic provisions of the NPP power unit commissioning programs and quality assurance programs in the course of the power unit commissioning shall be defined and substantiated, including division of the works into stages and sub-stages, their interrelation and mutual coordination, the procedure and time limits for completion of each stage or sub-stage, criteria of their successful completion, the required administrative and technical arrangements.

The following shall be demonstrated:

- the requirements of the GSR, NPP RF NSR and other regulatory documents are fulfilled to the full extent in the course of commissioning;

- safety is ensured in the course of commissioning works and testing at all stages of the NPP power unit commissioning;

- the required completeness of investigations and inspections is ensured for all modes and characteristics of the NPP power unit systems related to its safe operation assurance;

- the design basis and characteristics of the normal operation systems are confirmed.

13.2. Arrangement of works

The expected arrangement of the works and structure of interaction between the operating organization workers and representative of any scientific, design, engineering, installation, construction and commissioning organizations as well as the supplier organizations both in the course of preparation for commissioning and the NPP power unit commissioning shall be described.

Distribution of the managing and executive functions and responsibilities between the organizations involved in the works and between the executives at various levels aimed to achieve the commissioning goals and to solve any commissioning tasks shall be demonstrated. Arrangement of the works and selection of the workers in the organizations engaged in the works shall comply with the RD requirements.

The following shall be reflected in the presented information:

- the organizational structure of the operating organization including the NPP power unit workers, their rights and liabilities, requirements for qualification (the information shall be presented in case of any differences from the organizational structure provided within the commissioning period);

- the organizational measures implemented by the operating organization, the design developers, the equipment suppliers and other organizations engaged in performance of the works;

- functions of different organizations, their interaction and distribution of liabilities;

- plans for involvement of additional workers for each commissioning stage, the requirements for their qualification;

- description of the administrative measures for safety assurance, including radiation protection, fire safety, the relevant medical care, compliance with the sanitary and hygienic requirements, etc.

13.3. Stages of works

Division of the entire NPP power unit commissioning period into stages and sub-stages with due regard for peculiarities of the particular power unit and tasks to be solved at each stage (sub-stage) shall be substantiated, and information on the scope of the main commissioning stages shall be provided. In this case selection of the optimal sequence of the works, performance and (or) combination of tests, measures aimed to ensure proper control over their performance shall be explained, and the acceptance criteria shall be clearly defined.


- description of the NPP power unit commissioning schedule;

- commissioning works and acceptance tests for safety-related systems and safety systems;

- physical start-up and investigation of the neutron and physical characteristics of the reactor;

- power start-up and reaching the design NPP unit power.

Brief characteristics and the scope of works shall be presented for each stage and sub-stage of the commissioning works and tests, and peculiarities and purpose of the stages (sub-stages) shall be reflected; the way to perform the works with regard to the RP and auxiliary systems (including SS) shall be demonstrated.

13.4. Testing programs

Brief summary of the testing programs shall be provided for each stage (sub-stage) of the NPP power unit commissioning and information on the testing programs for all safety-related systems and for individual equipment shall be presented.

The following shall be specified for each stage (sub-stage):

- purpose of the works and tests, successful completion criteria;

- sequence of the works;

- requirements for readiness of the rooms, systems and equipment;

- process restrictions, conditions and arrangements for safe performance of the works and tests;

- composition, sequence, interrelation and duration of tests;

- the fundamental provisions of the work methodologies; in this case preparation for testing and testing methods for unique unparalleled equipment shall be described in more detail with indication of its acceptance criteria;

- requirements for the reporting documentation, particularly for its issuance, presentation and storage, as well as access to it;

- requirements for the number and qualification of the workers engaged in the works and tests and distribution of liabilities, including administrative services.

The planned way to use the information on the commissioning experience for similar NPP power units or NPPs with other reactor types shall be specified; it should be demonstrated how this information substantiates the relevant stages, methods and acceptance criteria in the newly developed program. Where possible the quantitative and qualitative parameters of the NPP power unit commissioning program shall be compared with the NPP power unit counterparts with regard to the scope, means, methodologies, methods for arrangement of the works and tests.

The stage, the way and the scope of the trial for normal, transient and emergency modes as well as methods and devices for the SS operation checking shall be specified. Specific and detailed information shall be provided to confirm that the planned works and tests will enable to comply with the above-mentioned safety conditions.

The following shall be described in detail:

- analysis procedures and methods used to reach the initial criticality and to measure the neutron and physical characteristics of the reactor nuclear core, particularly efficiency of the emergency protection for the nuclear core safety control;

- methods for assessment of the most important characteristics of the RP, SRS, SS equipment and the basic characteristics of the NPP power unit;

- special-purpose commissioning works on the systems with liquid-metal coolant and testing of individual safety-related NPP systems and equipment (for example, the reactor CPS, active and passive safety systems, refueling devices and hoisting mechanisms, etc.);

- potentially hazardous works and arrangements to prevent accidents.

The procedure for development and approval of the NPP power unit commissioning programs, quality assurance programs in the course of commissioning and working programs based on the design documentation shall be specified.

13.5. Schedule of works and testing

The schedule of the NPP power unit commissioning works with indication of the time limits for commencement of operation, delivery of the NPP power unit for commercial operation and the main stages shall be provided.

The main stages of the works, their estimated duration shall be indicated on the schedule; the list of all types of works and tests shall be presented separately for each stage. The planned adjustment and testing schedules for individual NPP facilities, systems and components shall be provided.

The schedules shall take into account the time for performance of the works as well as for processing, analysis and presentation of the results and their approval by the concerned parties in

accordance with the established procedure. The time necessary to develop more detailed or adjusted process operations or works at the NPP site and approve them prior to adoption, the time for development of detailed testing guidelines, emergency protection and operation manuals and training of the operating personnel shall be taken into consideration.

13.6. Additional requirements for the NPP power unit commissioning

Additional requirements to be taken into consideration in preparation for the works and in the course of the works at the NPP site shall be specified, particularly the requirements for:

- conditions for preparation, agreement and approval of the detailed design documentation (the operational technological procedure; the NPP SAR, the set of guidelines, particularly for actions in emergency conditions, etc.);

- participation of the operating and additional personnel in performance of the works and tests and development of the documentation, particularly the reporting one (including the requirements for the format of the reporting documentation);

- administrative and technical measures and actions in case of any beyond-design characteristics or deviations from the design, particularly the necessity to adjust the design and operation documentation;

- arrangement of maintenance and archiving of the documents;

- arrangement of the restricted access areas in the NPP power unit rooms and security zones depending on the phases and stages of the NPP power unit commissioning program;

- arrangement of fire-fighting services and control;

- arrangement of sanitary areas, radiochemical and radiometric control both in the rooms and near the NPP power unit;

- development and issuance of the certificate for industrial (commercial) operation of the NPP power unit;

- development and implementation of the emergency response plans and protection of the workers and the public in case of any accident at the NPP power unit.

13.7. CW completion report

Brief information on the results of testing stages shall be provided.

Performance of the planned works and fulfillment of the requirements as well as compliance of the characteristics of the facilities, systems and components with the design and effective regulatory documents shall be confirmed by documents based on the reporting materials and the results of the performed works and tests.

In case of any deviations from the design and effective regulatory documents the design documentation shall be adjusted with substantiation of any deviation acceptability with regard to the required safety and reliability level assurance presented in the relevant sections of the NPP SAR.

Any deviations from the performance and arrangement procedure occurring in the course of commissioning works and testing shall be described with analysis of the causes for such deviations and conclusions for future reference.

Compliance with the integrated work schedule of the NPP power unit commissioning program shall be demonstrated and analyzed from the viewpoint of completeness and time limits; feasibility of any deviations shall be assessed.

It should be stated what additional requirements for commissioning are fulfilled and to what extent of adequacy, including adjustment of the operation documentation in accordance with the results of the works.

14. REQUIREMENTS FOR THE SECTION "OPERATION"

14.1. Management arrangements

14.1.1. Operating organization

The scheme of the operating organizations and its departments performing the activities aimed to support the operation and the information on the principles and scheme of interaction between the NPP management and the operating organization shall be provided.

It should be demonstrated that the structure of the departments, distribution of liabilities and authorities among the departments, official duties of the workers, their qualification and responsibility enable the operating organization to perform the functions prescribed in the regulatory documents.

14.1.1.1. Structure of the operating organization and its departments

The list of the operating organization departments responsible for the following types of activities shall be provided in the block diagram:

1. The NPP design and construction

The operating organization departments (or any organizations engaged by the operating organization on the contractual basis) ensuring the following shall be listed:

- selection of the site with due regard for the natural and human-induced impacts;

- development of the designs of buildings, structures, the RP, safety systems and auxiliary systems;

- assessment of the design development level;

- compilation of the NPP SAR;

- supply of materials and equipment;

- performance of construction and installation works.

2. Pre-operational preparation

The list of departments responsible for implementation of any arrangements planned prior to the NPP commissioning commencement and submittal of the complete NPP SAR shall be presented. These arrangements include:

- development of the NPP commissioning programs;

- implementation of the NPP commissioning programs;

- development and implementation of the program for selection and training of the workers;

- development of the operation guidelines for the NPP commissioning and the operation manuals;

- determination of the initial radiation situation in the NPP location area;

- development of the annual equipment maintenance and repair plans.

3. Technical support of operation

The list of services responsible for arrangement of the operation support shall be provided:

- engineering and technical support of operation in solving of any problems related to nuclear and radiation safety and radiological protection;

- maintenance, repair and modifications of the thermomechanical and electrical equipment and mechanisms, instrumentation and controls;

- inspections and revisions including control of metal and weld joints;

- NF handling operations;

- quality (chemical composition) maintenance for the liquid-metal coolant of the primary circuit and the protective gas;

- RW management.

14.1.1.2. Organizational structure of the departments

Consultant Plus: note.There is probably a misprint in the official document test: there is no paragpah 13.1.1.1 in the document. It seems that the list specified in par. 14.1.1.1 is meant.

The structure of each department in accordance with the list specified in par. 13.1.1.1 with indication of the positions from the head of the department to the workers, the number of the personnel for each position with due regard for the reserve as well as the list of job descriptions shall be provided.

14.1.1.3. Qualification of the workers

The data on the positions providing more detailed information on the education level of the workers with indication of the education, training, specialties and professional experience in other positions and (or) organizations. Work of the persons without higher education occupying engineering positions (if any) shall be substantiated.

14.1.2. NPP administration and in-process management

The organizational structure scheme of the NPP power unit in-process management shall be provided.

The following shall be reflected in the presented information: the list of departments with their names and indication of the administrative management positions, structure of the departments, liabilities of the workers, their qualification and responsibility.

For the multi-unit NPPs the organizational scheme shall clearly reflect the planned changes and additions introduced to the organizational structure of the entire plant upon commissioning of the new

NPP power units. The schedule enabling to determine the time limits for occupation of all positions with commissioning of the new power units shall be provided.

14.1.2.1. In-process management organizational structure scheme

The following departments and services shall be specified on the block diagram:

- administrative;

- production;

- technical departments, laboratories and services.

14.1.2.2. Organizational structure of the departments

Consultant Plus: note.There is probably a misprint in the official document test: there is no paragpah 13.1.1.1 in the document. It seems that the list specified in par. 14.1.1.1 is meant.

The structure of each department in accordance with the list given in par. 13.1.2.1 shall be provided with indication of the positions from the department head to the workers (shift supervisors, shift operators, the repair personnel, etc.), the number of shifts as well as the number of the personnel for each position with due regard for the reserve (backup personnel).

The information on each NPP structural unit shall include:

- functions of the department;

- the procedure for interaction of the departments.

14.1.2.3. Rights and liabilities of the power plant workers

The list of job descriptions defining the rights and liabilities of the NPP power unit workers shall be provided. In particular, the procedure for succession of authorities (including transfer of the right to issue permanent and temporary directives and orders) and responsibility for the entire NPP operation for at least three officials (in case of any temporary circumstances) shall be described.

14.2. Workers

14.2.1. Requirements for the workers Qualification

Compliance with the RD provisions for selection of the workers to the positions specified on the structural schemes given in par. 14.1.1 and 14.1.2 in accordance with the required qualification (education, industrial experience, training), the requirements for psychological and physiological parameters and requirements for availability of the relevant permits issued by the State regulatory authority for safe atomic energy use shall be analyzed.

14.2.2. Training arrangement for the workers

Information shall be provided to demonstrate how the requirements of the GSR, NPP RF NSR, the NPU Rules and Basic Provisions for Selection, Training, Permit for Work and Control of the NPP Personnel in the Course of Operation and Selection of Workers for the Positions are implemented in training of the workers.

Analysis results for the training facilities and simulators for training of the workers as well as compensation measures in the absence of any full-scale simulator of the NPP power unit or its inconsistency with the particular unit shall be provided.

14.2.3. Coordination (correlation of stages) of the personnel training with the CW stages and NF loading Staff recruitment schedule

The schedule of each stage of the operating personnel training for performance of the NPP power unit commissioning stages shall be provided (or references to Section 13 shall be given), and the expected time limits for NF loading and staff recruitment for the NPP power unit shall be specified.

The necessary time limits for admittance of the workers to the workplaces, presence of the workers admitted to adjustment of the equipment and systems and workers of any other organizations directly engaged in the commissioning works, physical start-up and power testing shall be presented in the schedule.

14.2.4. Qualification maintenance for the workers

The system of the qualification level monitoring for the workers and the arrangements aimed to maintain the required qualification including regular exercises and traning on the simulators in order to drill the actions under normal operation conditions and in emergency situations shall be provided. The way to ensure compliance with the GSR requirements for training of the workers and analysis of any errors shall be reflected.

14.3. Guidelines

14.3.1. Development of guidelines

The NPP power unit operation stages when the relevant guidelines are to be developed and put into effect shall be specified.

14.3.2. Job descriptions

Information on the job descriptions for the administrative, management and operating personnel shall contain their list in accordance with the operating organization structure and organizational arrangement.

14.3.3. Operation manuals

14.3.3.1. Process regulations

The basic fundamental provisions of the process regulations shall be presented.

14.3.3.2. Operation manuals for the equipment and systems

The list of operation manuals for the power plant equipment and systems shall be provided, the procedure for the operating personnel to find the relevant guidelines on actions in case of any alarm signals and identification of any initiating events for arising emergency situations shall be specified, and the manuals the operating personnel shall know to the full extent shall be listed.

14.3.3.3. Maintenance and repair guidelines

The lists of in-plant, factory and standard guidelines to be used in the course of maintenance and repair of the main and auxiliary equipment of the systems, inspection of the protections, automatic devices and other systems specified in the relevant sections of the NPP SAR shall be presented.

14.3.3.4. Safety guidelines

The list of safety guidelines to be available at each workplace together with the operation manuals in accordance with the list of technical documentation for each workplace approved by the chief engineer (the director) or at the workplace of the department head shall be presented.

14.3.3.5. Guidelines for maintenance of operational documentation

The information related to the guidelines for maintenance and handling of operational documentation shall contain the prescribed procedure for maintenance of the operational documentation by the duty personnel, the place of its permanent storage, the requirements for the documentation integrity and the period of its storage.

The actions of the administrative and technical workers of the power plant aimed to control maintenance of the operational documentation shall be described.

14.3.4. Emergency instructions

14.3.4.1. The list of emergency instructions shall be provided:

- instructions on elimination of any operational occurrences and emergency situations;

- instructions on elimination of design basis accidents;

- instructions (guidelines) on beyond design basis accident management.

14.3.4.2. The requirements specified in the manuals shall contain:

- actions of the workers aimed to identify emergency situations and accidents;

- the required number of the operating personnel (with indication of specific positions) for performance of any corrective actions;

- characteristic features of success (failure) in performance of any actions with the equipment;

- criteria for proceeding to the actions in accordance with the accident management guidelines.

14.3.5. Accident management guidelines

Brief information on the accident management guidelines shall be provided. It may be presented in a separate appendix to this section.

14.4. Maintenance and repair

14.4.1. Annual equipment maintenance and repair plans

The annual plans of the equipment maintenance and SPM with indication of the main types and scope of activities (general maintenance, overhaul repair, repair and replacement of components, testing, modification of the systems, etc.) shall be provided.

The way to ensure efficient and timely assistance of the design organization in case of any

malfunction and the necessity to modify individual assemblies shall be demonstrated.

The schedule of preventive maintenance shall be presented.

14.4.2. Maintenance conditions

The list of the following maintenance means shall be provided:

- instrumentation;

- means for decontamination and maintenance of radioactive assemblies;

- lifting and handling equipment;

- special equipment and tools.

Availability of the means, materials, spare parts, etc. as well as the workshops for repair of the equipment shall be demonstrated.

14.5. Arrangement of control and presentation of information on the NPP power unit operational safety level

Information on the adopted system for monitoring of the operational (current) state of the NPP power unit, the procedure for data collection and analysis as well as presentation of information related to safety shall be specified.

14.5.1. Control by the operating organization representatives

Information on the measures planned by the operating organization to perform inspections of the NPP power unit for compliance with the RD requirements shall be provided.

14.5.1.1. Inspection program

The planned inspection program shall be provided with indication of the following:

1. Type of inspection

2. Scope of the inspection with regard to the following basic issues:

- compliance with the requirements of operation manuals and state of the operation documentation;

- quality assessment of the chemical composition maintenance for the liquid-metal coolant and protective gas and the metal state monitoring for the equipment;

- checking of the state of systems and equipment;

- inspection of nuclear and radiation safety;

- checking of the system for selection, training, admittance to unsupervised works and qualification maintenance for the NPP workers, checking for compliance with the emergency response drilling procedure;

- checking of fire safety and other emergency prevention arrangements;

- performance of repair and preventive works;

- compliance with the requirements of the state safety regulating authorities;

- assessment of the NPP QAP functioning in the course of operation;

- assessment of the safety culture state.

3. Frequency of inspections

4. Criteria for assessment of the inspection results enabling to define whether the power plant is operated in compliance with the regulatory requirements and QAP for operation (Chapter 17).

5. Procedure for presentation of the inspection results as well as the requirements for storage and access to the reporting documentation

14.5.1.2. Organizational structure

Information on the operating organization departments and officials implementing the in-plant inspection program, their number and qualification shall be provided.

14.5.2. Preparation and presentation of periodic information on the current safety level

The information shall comply with the requirements of the effective provisions related to annual reports on assessment of the current operational safety state at the power units and the procedure for investigation and recording of any operational occurrences at the NPP power unit.

14.6. Physical protection

The main administrative and technical arrangements aimed to prevent any unauthorized actions of the workers or any other persons in relation to nuclear materials, radioactive substances and RW or any safety-related NPP power unit systems, equipment and devices capable to cause (directly or indirectly) any accident and pose any hazard for the health of the power plant workers and the public due to radiation impact shall be demonstrated. The information provided in this section shall confirm compliance with the requirements of the Physical Protection Rules for Nuclear Materials, Nuclear Installations and Nuclear Material Storage Facilities, as well as any other effective regulatory documents and industry-specific administrative documents.

14.6.1. Configuration of physical protection and requirements thereof

The following shall be defined:

1. Engineering and technical sub-systems with the description of:

- the security alarm systems;

- the access control systems;

- the closed circuit television systems;

- the in-process communication systems;

- the engineering security features;

- auxiliary systems and means supporting functioning of the physical protection.

2. Administrative measures (in the form of sub-system), namely:

- arrangement of the NPP security including training of the security workers;

- preparation of the NPP workers for actions in emergency situations;

- arrangement of access to the protected area and the vital areas for the permanent and shift NPP workers;

- arrangement of the NM accounting, storage, usage, protection, transportation and control system;

- arrangement of personal and special inspection of the workers, secondees, visitors and vehicles, etc.

3. It should be specified that the PPS is referred to safety systems and shall be designed with due regard for the following requirements:

- independence;

- multiple channels;

- fire safety;

- operability and reliability under the design internal and external impacts of natural and human-induced character.

14.6.2. PPS schemes and structural arrangement

The basic principal schemes of the engineering features for PPS monitoring and alarms shall be provided.

The principal structural arrangement of the PPS with regard to the security organization shall be presented without disclosure of the location of control panels, signal and observation stations.

Access to the PPS materials shall be restricted due to the fact that these materials are confidential.

14.7. Emergency planning

Information on the planned arrangements for protection of the workers and the public in case of any accidents in accordance with the requirements of the GSR, the Action plan for protection of the workers and the public in case of any radiation accident at the NPP, the Provision on the procedure for the emergency situation announcement, in-process transmission of information and arrangement of urgent assistance to nuclear power plants in case of any radiation-hazardous situations, Construction Rules and Regulations (Civil Defense Engineering and technical Measures) and any other regulatory documents for protection of the workers and the public shall be provided.

14.7.1. Protection of the workers

Information shall provide clear concept of the planned and implemented measures for protection of the workers in case of any accident at the NPP and shall reflect:

1. Emergency preparedness and intervention levels

2. Administrative measures in case of accidents, including:

- distribution of liabilities and coordination of actions with any external organizations within the

NPP site and SPA (the fire-fighting service, medical institutions, local authorities);

- actions of the officials ensuring announcement of accidents and commencement of the personnel protection plan implementation in case of any radiation accidents at the NPP and the NPP power unit;

- conditions and communication means used for the announcement.

3. Types of accidents that can occur at the NPP or are considered in the emergency action plans and the ways to inform the workers.

4. Types and amounts of radioactive substances that can be released into the NPP rooms, the radiation exposure paths and protective means.

5. Time of access and stay of the people in specific NPP areas (this is referred in particular to the control rooms and emergency response control posts).

6. Instrumentation necessary in case of accidents (their suitability for fast detection and continuous assessment of the radiation situation in the course of accidents, their functional capabilities, including the measurement range and response time; location of the sensors and recording equipment; availability of spare and redundant instruments; emergency alarms).

7. Number of the personnel and means required to assess the situation, to implement the protective measures, to arrange communication and to maintain the reporting documentation as well as to render any help to the injured.

8. Criteria used to commence evacuation of the workers, marking of the evacuation routes, indication of the areas for gathering of the NPP workers, first aid treatment and estimation of the required medicines.

9. Availability of any protected emergency response posts at the NPP and in the city equipped with computers, communication means, annunciation means, means for collection of information on the radiological and meteorological situation at the NPP territory, in the SPA and the supervised area.

10. Availability of shelters complying with the requirements of civil defense standards for accommodation of all NPP workers, workers and official of any other organizations (including the staff of military units and fire brigades) supporting functioning and activities of the NPP.

11. Preparedness of the local annunciation systems for the NPP workers and the public within the area of 5 kilometers.

12. Readiness of the residential and industrial buildings and facilities at the NPP territory and in the city near the NPP for initial sheltering of the workers and their family members (in case of insufficient number of shelters).

13. Planning of the arrangements for preparation of the main and backup evacuation areas for acceptance of the workers and their family members in case of any accident at the NPP power unit.

14. Availability of sufficient number of special-purpose cars, vans and buses with leak-tight cabins equipped with removable filtration and ventilation units and intended for delivery of food and transportation of the service personnel in case of any radiation accidents at the NPP.

15. Presence of the developed measures for protection and usage of water resources within the SPA and the supervised area.

16. Administrative measures in case of any emergency situation including the procedure for coordination of the NPP personnel actions with the facility-based and territorial forces of the Russian

EMERCOM, civil defense services, local authorities, ministries and agencies engaged in protection of the public and mitigation of the accident.

14.7.2. Emergency response control posts at the NPP

Information on the emergency response control posts at the NPP as well as in any places where they are unlikely to be affected by the accident simultaneously with the control posts at the site shall be provided.

In this case the following shall be specified:

- the location of the post selected in such a way so that free movement towards or from it would not be severely obstructed in any emergency situation;

- the staff of the control post and their qualification;

- the list of equipment in the post as well as its storage conditions and readiness maintenance (it should be demonstrated that the technical means of the emergency response control posts, the instrumentation pool, communication, personal protection equipment, etc. are operable and perform their functions properly in any emergency situations).

14.7.3. Mitigation of accidents

Potential consequences of accidents and the relevant mitigation arrangements shall be specified; the decontamination methods and means for the main and auxiliary equipment, facilities and the area, the methods and means to render help to irradiated workers and the public including information on sanitary treatment and medical care, the list of medicines, dressing supplies and other auxiliary materials with indication of their storage places, the decontamination methods and means for the radioactive contamination zones shall be also described.

14.7.4. Emergency response training

Information on the programs and schedules (at the SAR stage) of the emergency response training and drilling shall be provided with indication of the categories of administrative staff and workers engaged in exercising of the relevant actions in the course of accidents and mitigation of accidents, as well as the applied engineering features (including simulators) for the exercises and the reference time standards for performance of the actions.

15. REQUIREMENTS FOR THE SECTION "ANALYSIS OF THE NPP POWER UNIT OPERATIONAL OCCURRENCES INCLUDING ACCIDENTS"

15.1. Analysis of emergency transient processes

The NPP safety assessment shall include analysis of the response of the NPP systems and structures to any potential initiating events.

This analysis shall constitute an integral part of the NPP safety analysis.

The scenarios of the expected events and their consequences shall be defined, and the possibility for intervention into the operation of the systems in order to control the progress of the processes shall be also assessed.

This analysis shall be used as the basis for arrangement of the control for NPP systems in different

situations.

15.1.1. The list of initiating events for operational occurrences

The approximate list of initiating events:

- loss of integrity (leakage) of the reactor pressure vessel;

- leakage of the main primary circuit pipeline (the pressure pipeline in the reactor pressure vessel with integral concept);

- the RCP shutdown in different modes;

- erroneous voltage supply to the RCP motors with their switch-on to the nominal rotation rate with the CPS rods disenagaged (in the refueling mode);

- closure of a check valve during operation of all primary circuit RCPs;

- erroneous opening of the check valve in the non-operated loop during the reactor operation with other heat removal loops;

- intercircuit IHE leakage;

- unintended movement of a control rods in various reactor states;

- unintended movement of a shim rod in various reactor states;

- unintended lifting of an EP rod in the course of the reactor start-up;

- falling of a fuel assembly into the reactor in the course of refueling;

- ingress of hydrogen-contaning substances into the nuclear core;

- occurrence of gas bubbles in the nuclear core and their passing through fuel assemblies;

- deterioration of heat removal from fuel assemblies in the spent fuel assembly drum;

- water leakage to sodium;

- loss of feedwater supply to one or all SGs;

- loss of the system power supply (loss of auxiliary power supply);

- fire in the NPP power unit rooms;

- failure of non-tank ionizing chambers particularly due to malfunction of the ionizing chamber unit cooling;

- the secondary circuit RCP shutdown;

- disconnection of the turbine generator from the grid;

- the turbine shutdown;

- leakage of the main steam pipeline;

- falling of heavy objects on the reactor;

- seismic impacts;

- shock waves;

- floods;

- loss of cooling water;

- tornado.

The list of initiating events may be changed based on the analysis of the particular RP scheme, operation modes and maintenance regulations.

15.1.2. Analysis of emergency transient processes

The analysis materials for each initiating events shall be presented in the following sequence.

15.1.2.1. Initial state of the NPP power unit and its systems before the initiating event

The state of the NPP power unit systems and components at the moment of the operational occurrence onset shall be described in detail for each initiating event. In this case the particularization degree in the IE description shall depend on the nature of the operational occurrence. This description shall be sufficient for the further safety analysis. The reactor power level, background of its operation determining the state of fuel elements, fuel burn-up in these elements and mechanical properties of the materials, coolant flow rates in the circuits (in case they can change due to changes in the circulation pump rotation speed); parameters of the third circuit coolant, temperature of the coolant in the basic points of the primary and secondary circuit; positions of the absorbing rods; gas pressure in the reactor and the SG buffer tank, sodium levels in the reactor and the circulation pumps shall be specified.

In case of necessity power density distribution in the nuclear core shall be presented in the coordinate system, particularly power density due to fission and induced activity products. In case the analysis is performed for the mode with any deviations from the steady one the description of the above-mentioned parameters shall be provided in this section of the report. Thus, if the initiating event is due to water leakage to sodium in the SG the distribution of sodium, water and steam flow rates among the SG sections, the distribution of temperatures in the sections, hydrogen content in sodium and protective gas and any other parameters important for investigation of non-steady processes at the NPP power unit shall be described. Likewise, detailed description of the initial state of the spent fuel assembly drum and the fuel assemblies contained in it shall be required in the analysis of any malfunctions of FA cooling in the spent fuel assembly drum. Particularization of the initial state description for each IE shall be defined by the NPP SAR developers. The IE consequences can depend significantly on the initial power of the reactor. In this case the maximum reactor power does not always correspond to the most severe consequences of an operational occurrence (for example, in case of excess reactivity due to movement of absorbing rods or ingress of any moderator materials into the nuclear core). In case of some CSS failures consequences of the excess reactivity at very low reactor power levels can be more severe than at high power levels. In case it is difficult to assess the most unfavorable initial state of the NPP power unit from the viewpoint of the IE consequences the analysis shall be performed for different initial conditions covering all possible variants of processes in the RP and its systems.

The list of input data necessary and sufficient for check calculations shall be given in Appendix 5.

15.1.2.2. Functioning of the systems

Functioning of all systems intended to support operation of the NPP power unit without any deviations from the safe operation limits shall be described. Description of the operational occurrence

shall be presented, the required values and change rates for the input parameters of the reactor determined by the properties of safety-related systems shall be specified.

15.1.2.3. Consideration of potential system failures

Potential failures of safety-related systems shall be considered for each initiating event. Possible failures shall be taken into account in description of the SS functioning in accordance with the design algorithm. In compliance with the NPP RF NSR failure of the most efficient protection device shall be considered as one of the failures in all cases requiring activation of the reactor EP.

15.1.2.4. Analysis methodology

Mathematical models and computational programs used for calculations of the non-steady processes in case of any operational occurrences shall be described. In case any experimental data are applied in the analysis the conditions for obtaining of these data shall be described in brief, and the possibility to use them in the case under consideration shall be substantiated. References to the sources where these data are published shall be given. The particularization degree in description of the mathematical models and computational programs shall depend on their validation state. Brief description explaining the essence of the applied models and assumptions with the reference to the relevant validation documents shall be sufficient for validated programs. For non-validated programs the description shall be detailed, and information on the mathematical models, assumptions, solving methods, verification of the programs and comparison of the calculations with the experimental data (if any) shall be provided. Any updates of the validated programs shall be described and substantiated in case such programs are used to analyze non-steady processes.

15.1.2.5. Input data for the analysis

The input data required for analysis of non-steady processes at the NPP power unit (the constructive characteristics of the systems, parameters characterizing their operation modes, neutron and physical characteristics of the nuclear core and its dynamic characteristics, including reactivity feedback), thermal, physical and mechanical properties of the materials, etc. shall be described. The complete set of the input data shall be defined with due regard for the operation of the NPP power unit components where the main changes characterizing the operational occurrence consequences take place. In case the input data for the analysis are contained in any other sections of the NPP SAR the reference shall be given in the section with indication of the number of section, table, figure where these input data are presented. In any other cases when any data beyond the framework of the description in the NPP SAR sections are used for calculations these data shall be presented in this section with indication of the source they are taken from.

15.1.2.6. Results of analysis

The main purpose of the RP operational occurrence analysis is to substantiate the design requirements for fast response, efficiency and other characteristics of safety systems and to confirm compliance with the safety criteria and requirements in the NPP power unit design. The control signal generation schemes, the response levels of the alarm devices for the relevant parameters and permissible delays of signal generation shall be substantiated for CSSs of the NPP power unit. These characteristics shall be analyzed taking into account operation of the safety-related normal operation systems determining the reactor disturbance rate and value with regard to the input parameters.

Analysis of the operational occurrences caused by loss of power supply for the circulation pumps, malfunctions in the feedwater supply to the SG shall enable to substantiate the configuration, structure and characteristics of the nuclear core cooling system and the system for heat removal to the ultimate heat sink.

Analysis of any near-accident situations shall demonstrate efficiency of the safety-related systems

provided in the design and impossibility for any near-accident situation to develop into an accident.

15.1.2.7. Assessment criteria

The main SS efficiency criterion in the analysis of operational occurrences is non-exceedance of the safe operation limits. The following shall be taken into account in the assessment of fuel element damages:

- the design number of modes (substantiated by the expected IE frequency and the probability of postulated SS failures) caused by the operational occurrence;

- deformation of the nuclear core components due to thermal, mechanical and radiation impacts;

- physical and chemical interaction of the nuclear core materials;

- limit values for the thermotechnical parameters of the nuclear core;

- vibrations, thermal cycles, fatigue and ageing of the materials;

- impact of fission products and the coolant impurities on corrosion of the fuel element claddings;

- impact of radiation and any other factors deteriorating mechanical characteristics of the nuclear core materials and integrity of the fuel element claddings.

Apart from the fuel elements, the primary circuit damages shall be taken into account in assessment of the IE consequences. In accordance with the NPP RF NSR it should be demonstrated that the primary circuit systems and components operate reliably within the design service life with due regard for corrosive, chemical, thermal, force and other impacts possible under normal operation conditions and in case of any operational occurrences and design basis accidents. It is specified in the NPP RF NSR that the basic design of the RP shall demonstrate that the reactor pressure vessel strength in accordance with the strength standards is ensured under normal operation conditions, in case of any operational occurrences and design basis accidents within the entire RP service life. The above-mentioned criteria shall be used as the basis for assessment of near-accident situations.

15.1.3. Conclusions

The results of the analysis shall be provided, and the conclusion on compliance of the design with the RD requirements for safety and fulfillment of all criteria specified in these documents shall be made.

15.2. Analysis of design basis accidents

15.2.1. The list of initiating events for design basis accidents

The list of initiating events for design basis accidents at the NPP power unit with fast reactors with liquid-metal sodium coolant (to be considered in the safety analysis):

- flow passage reduction or blockage in one fuel assembly due to swelling of the materials, ingress of coolant impurities or foreign objects with subsequent breakage and melting of fuel elements;

- loss of leak-tightness in the primary circuit pipeline at any section without safety containment;

- loss of leak-tightness in the primary circuit gas system;

- loss of integrity in the spent fuel assembly drum.


Operation of the systems and components in case of accidents shall be described. The analysis results for each design basis accident shall be presented in the following sequence.

15.2.2.1. Initiating event

The requirements for description of initiating events for the NPP power unit systems and components before the design basis accident and scope of this description as well as any accident development paths shall be supplemented with assessment of the radiological consequences of design basis accidents. As the accident consequences include loss of integrity in fuel element claddings, sodium pipelines, the primary circuit gas system the IE description shall include the information enabling to define the amount and nuclide composition of radioactive substances in sodium and protective gas of the primary circuit, the amount and nuclide composition of fission products in fuel elements. The state of the primary circuit rooms where the hazardous components (sodium pipelines, gas systems) are located shall be described. The gas pressure and temperature in these rooms (in accordance with the design) and their leak-tightness degree required for subsequent assessment of RSb propagation into the NPP power unit rooms shall be also set.

15.2.2.2. Functioning of safety systems in accordance with the design algorithm after the accident onset

The design activation sequence for CSSs, PSSs and LSSs ensuring non-exceedance of the RP safe operation limits shall be described.

15.2.2.3. Consideration of potential failures of safety systems and human errors in analysis of design basis accidents

In accordance with the requirements of the GSR and NPP RF NSR the list of postulated SS failures to be considered in analysis of design basis accidents shall be provided. The basic postulating principles for these failures shall be assumed the same as for the analysis of operational occurrences.

15.2.2.4. Analysis methodology for accident processes

Mathematical models and computational programs used for calculations and analysis of design basis accidents shall be described. Models and programs describing not only accident processes in the reactor nuclear core, the spent fuel assembly drum, the sodium burning processes in the primary circuit rooms and leakages of radioactive gas from the gas system but also propagation of radioactive substances into the NPP power unit rooms shall be included. Brief description with the reference to the relevant documents shall be provided for validated computational programs. Non-validated programs shall be described in more detail. Special attention shall be paid to verification of these programs. Brief description of the calculation and experimental data confirming the adequate level of accuracy for the applied programs, as well as references to the sources where they are published shall be provided.


The complete set of input data required for calculations of accidents and analysis of their consequences shall be described. References may be given to any other sections of the NPP SAR where the constructive characteristics of the investigated facility, the operation mode descriptions, the rated or any other parameters (power levels, temperatures, coolant flow rates, pressure, etc.) are presented. Special attention shall be paid to non-standard parameters: thermal, physical and mechanical properties of the materials in the high temperature area, equations for the state of materials in the emergency temperature and pressure deviation area. In all cases references to the sources these data are taken from shall be given.

15.2.2.6. Results of design basis accident analysis

Results of calculations and subsequent analysis of the accident processes in case of design functioning of safety systems, any system failures and human errors postulated in accordance with the RD requirements for safety shall be provided. Analysis results for the accident processes shall be used as substantiation for the SS characteristics stipulated in the design.

It should be demonstrated that no deviations from the safe operation limits with regard to damage of fuel elements, damage of the primary circuit, excessive radiation exposure for the workers and the public will occur. The calculation results for the accident processes in the reactor nuclear core, the spent fuel assembly drum, the primary circuit rooms as well as the calculation results for releases of radioactive substances outside the primary circuit shall be provided.

15.2.2.7. Radiological consequences of design basis accidents, calculation of equivalent exposure doses for the workers and the public

Calculation results for propagation of radioactive substances into the NPP power unit rooms and outside them shall be presented. The calculation shall be performed with due regard for the data on leak-tightness of the primary circuit rooms and the worst weather conditions. Equivalent and effective exposure doses for the workers and the public after an accident shall be determined. Recommendations for the workers with regard to actions in emergency conditions shall be provided and reflected in the relevant guidelines. Accordingly, the necessity for any protective measures for the public living in the supervised area shall be considered.

15.2.2.8. NPP power unit safety assessment criteria in case of design basis accident

The criteria for assessment of the consequences for the particular accident under consideration from the viewpoint of radiation safety shall be specified. In case of accidents with the risk of fuel element damages, increase of temperature and pressure in the primary circuit the safety criteria shall be selected based on the requirements for non-exceedance of the maximum design fuel element damage limit, the safe operation limit with regard to fuel element damage (NPP RF NSR) except for any damages of the primary circuit and the reactor vessel.

Besides, the limitations of the radiation exposure for the NPP power unit workers and the public after an accident established in the RSS shall be observed for design basis accidents.

15.2.3. Conclusions

Results of the design basis accident analysis shall be provided. Consequences of accidents shall be described in brief, the conclusion on the NPP power unit safety assurance in case of these accidents shall be made based on the criteria specified in par. 15.2.2.8. Special attention shall be paid to radiological consequences of accidents and compliance with the RSS requirements.

15.3. Analysis of beyond design basis accidents

15.3.1. List of beyond design basis accidents and its substantiation

15.3.1.1. Groups of beyond design basis accidents

The list of beyond design basis accidents shall include the accidents capable to result in melting of the nuclear core and ultimate release of radioactive substances into the environment. The selected accidents shall be divided into the following groups:

- accidents caused by reactivity changes;

- accidents caused by any malfunctions of the nuclear core cooling;

- accidents caused by any loss of fuel cooling on the refueling route;

- accidents caused by leakage of radioactive sodium or gas from the primary circuit;

- accidents caused by mechanical damages of fuel assemblies in the course of refueling.

The accidents selected for analysis shall pose potential hazard of NF damages or dangerous releases of radioactive substances outside the primary circuit.

The list of beyond design basis accidents shall be substantiated based on the analysis of PSA results.

Compliance with the design criteria shall be confirmed in the course of accident analysis.

15.3.1.2. Scenarios of beyond design basis accidents

All scenarios of beyond design basis accidents resulting in exceedance of exposure doses for the workers and the public and the norms of RSb releases and content in the environment established for design basis accidents shall be specified based on the analysis results. The vulnerable points of the NPP shall be defined through the minimal cut sets of event (failure) trees. This term shall hereinafter mean combinations of the NPP design peculiarities, schematic solutions, layout, operational procedures and the organizational structure of any activities of the workers representing the most probable causes of the reactor nuclear core damage beyond the damage limits permissible for design basis accidents.

15.3.1.3. Characteristics groupd of beyond design basis accident scenarios

The scenarios specified in par. 15.2.1 shall be united into groups; the "response" of the plant systems required to prevent the accident development shall be the same within each group (the system functional event trees developed in the PSA section shall be similar).

15.3.1.4. Representative scenarios of beyond design basis accidents

One or several representative scenarios complying with the following four criteria in total shall be defined within each group (par. 15.3.1.1):

1. The maximum dose rates for the workers and (or) the public

2. The maximum intensity of radionuclide releases

3. The maximum integral radionuclide release

4. The maximum level of damage for the NPP systems and equipment

15.3.1.5. List of beyond design basis accidents

The scenarios defined in par. 15.3.1.4 shall be summarized in the list of beyond design basis accidents for further analysis.

15.3.2. Sequence of beyond design basis accident analysis

Each beyond design basis accident shall be analyzed in the following sequence.

15.3.2.1. Initial state of the NPP power unit before the accident

Requirements for description of the initial NPP power unit state before the accident shall be

similar to the requirements for description of design basis accidents.

15.3.2.2. Analysis methodology

Mathematical models and computational programs used for analysis of the corresponding beyond design basis accident, assumptions and uncertainties applied in the calculation methodologies, experimental data (if any) shall be described. As the analysis of beyond design basis accidents requires modelling of complex processes in the reactor with changes of the physical state of the nuclear core materials, complicated spatial and temporal nature of the heat and mass transfer processes inside the reactor the information on verification of the relevant programs shall be of particular value. This information shall be presented in brief with references to the relevant publications. The ST validation state shall be also specified, and the possibility to apply the program for analysis of the corresponding accident shall be substantiated.


Characteristics of the NPP power unit systems and components enabling to model the processes in the facility under consideration shall be provided. Description of the NPP site and the surrounding area, hydro- and meteorological data, information on distribution of populated localities in the NPP region required for subsequent calculation of propagation of any radioactive products released from the reactor in case of an accident within the surrounding area, as well as equivalent effective exposure doses for the workers and the public shall be additionally provided.

15.3.2.4. Calculation results for accident processes, assessment of RSb releases from the primary circuit in case of accidents

Calculation results for accident processes in the NPP power unit reactor and any other facilities under investigation shall be described in accordance with the beyond design basis accident scenario. The description shall contain sufficient details and include the information on spatial and temporal distribution of the most important parameters of accident processes; information on the existing margins before the accident process transition to the next critical phase corresponding to the certain level of the nuclear core damages and releases of radioactive substances outside the primary circuit shall be provided. The beyond design basis accident calculation shall be completed with determination of such release. Calculation results for releases of radioactive substances outside the primary circuit shall be further used to calculate propagation of radioactive substances in the NPP power unit rooms and in the environment. Propagation of gaseous, volatile and aerosol radioactive substances, their deposition on the surfaces of the rooms and the LSS filters shall be taken into account in the calculations. The most unfavorable of all possible data on leak-tightness of the industrial rooms and the weather conditions shall be assumed for calculations. All potential exposure paths for the public (direct exposure from the drifting cloud, from the cloud footprint, inhalation exposure, ingress of radioactive substances to the human organism via food chains) shall be taken into consideration. Conclusion on compliance with the RSS requirements and the necessity for any protective measures including evacuation of the public shall be made based on the calculation of effective and equivalent exposure doses for the workers and the public within one year after the accident.

15.3.2.5. Beyond design basis accident management

- Immediate safety objectives

The immediate safety objectives shall be defined for each beyond design basis accident severity level, i.e. the objectives the NPP operating personnel shall strive to achieve under these conditions in order to prevent or stop further development of the equipment and (or) SRS damage or to limit releases of radioactive substances into the environment.

- Facility condition characteristics, criteria of the beyond design basis occurrence and development

The facility condition characteristics shall be defined based on the performed calculation analyses of beyond design basis accidents, and the criteria that can be used (together with the condition characteristics) to determine the fact of the beyond design basis accident occurrence and to trace development of the relevant accident severity levels.

- Systems and equipment that can be used to achieve the safety objectives and to mitigate the accident consequences

All technical systems of the NPP (including the systems not related to safety assurance) that may be used (perhaps for the purposes other than the design ones and not in the design operation modes) in order to achieve the immediate safety objectives and to mitigate the accident consequences at each severity levels shall be defined. Issues of redundancy shall be considered for the systems performing the same function. The possibility to use materials and equipment located at the neighboring power units as well as outside the NPP site shall be described, and the means for their delivery shall be specified.

- Success criteria for the actions

Success criteria for the actions of the workers aimed to achieve the immediate safety objectives at each accident severity level shall be defined. Expression of these criteria via the condition characteristics shall be determined.

- Analysis of the scope of information on the facility condition available to the operating personnel in the course of accident development

The scope of information required to monitor the facility condition characteristics, to establish the accident severity levels, to control the necessary technical systems and to assess success of any actions for beyond design basis accident management as well as any hardware and techniques enabling to obtain this information under the expected conditions shall be defined. In case of necessity to perform indirect assessment of the required parameters the methods of such assessment shall be provided.

- Strategy of corrective actions

The strategy of corrective actions of the workers in case of any beyond design basis accident aimed to achieve the safety objectives at all potential accident severity levels shall be described.

15.3.3. Information on assessment of any probabilities of the nuclear core damage and hazardous releases of radioactive substances based on the results of beyond design basis accident analysis

Probability of any nuclear core damage and hazardous releases of radioactive substances shall be assessed.

The entire complex of the obtained data shall be reviewed and described, and preliminary conclusions on the potential ways of the nuclear core development and hazardous RSb releases shall be made.

15.3.4. Conclusions

Results of the beyond design basis accident analysis and the conclusion on compliance with the RD requirements shall be provided.

16. REQUIREMENTS FOR THE SECTION "SAFE OPERATION LIMITS AND CONDITIONS. OPERATION LIMITS AND CONDITIONS"

16.1. Safe operation limits and conditions

16.1.1. Safe operation limits

The safe operation limits, the controlled parameters, the technique and the exact point of their measurement, substantiation of the adopted permissible limits in accordance with the safe operation conditions and their measurement accuracy, the parameter change and measurement ranges, accuracy of the performed calculation and (or) experimental substantiation of the parameter (it is recommended to give references to Sections 4 and 15), the permissible information loss period, redundancy of the measurement channels shall be provided.

In general, information on the following safe operation limits shall be presented:

- NF temperature;

- fuel element cladding temperature;

- the reactor power level and its increase rate;

- temperature of the primary and secondary circuit coolant, the working media of the third circuit;

- the coolant level in the reactor tank, the tanks of the primary and secondary circuit pumps;

- flow rate of the primary circuit coolant;

- rotation rate of the primary and secondary circuit pumps;

- gas pressure in the reactor;

- number of leaky fuel elements;

- the NF burn-up depth and power of fuel assemblies;

- activity of the primary and secondary circuit coolant, the primary circuit protective gas, air (or gas) cooling the biological protection and the reactor cavity, the coolant of the spent fuel assembly drum, the water in the spent fuel assembly pool, releases and discharges of radioactive substances into the environment;

- contamination of the coolant with any foreign impurities;

- permissible number of different thermal cycles for safety-related components and equipment;

- permissible level of potential mechanical, thermal and other impacts on the NPP power unit systems and equipment caused by design basis accidents and external factors;

- the limit parameters (temperature, pressure) of the reactor pressure vessel, the safety containment, the secondary and third circuit.

16.1.2. Setpoints for activation of safety systems

All setpoints for activation of safety systems shall be specified. The adopted setpoint values shall be substantiated, the modes (processes) that define reaching of setpoints as well as their measurement accuracy shall be specified. Actuation setpoints of the warning and emergency alarms shall be specified and justified with substantiation of the interval to the SS actuation setpoint values.

It should be demonstrated that actuation of safety systems ensures non-exceedance of the safe

operation limits with due regard for the lag effect. The existing margins shall be provided.

The list of conditions when the operator shall shut the reactor down shall be provided.

In case the design allows activation of safety systems by the operator the following information shall be presented:

- that the operator is provided with the relevant information prepared in accordance with the requirement for CSSs (description of this information support shall be presented in Section 12, subsection 12.5);

- that the operator has enough time to initiate the safety systems, and control of the commands is arranged in such a way so that the operator's commands for activation of safety systems have the same priority as the CSS commands;

- consequences of any operator's errors shall be analyzed.

16.2. Safe operation conditions

16.2.1. Power levels and permissible normal operation modes

The permissible normal operation modes (for example, operation at the partial power level, operation with incomplete number of loops, warm-up and cooldown modes, refueling, etc.) and the corresponding permissible power levels shall be provided, definitions of the above-mentioned modes shall be presented.

The operation limits of the basic parameters such as power, protective gas pressure, temperature of the coolant, the temperature changing rate, chemical composition and radioactivity of the primary circuit coolant, the reactivity margin shall be specified for the permissble normal operation modes and each power level.

The avove-mentioned limits shall be expressed via the parameter values controlled by the operator, otherwise the relation between the limiting parameter and directly controlled parameters shall be demonstrated through the use of relevant tables, diagrams or calculation methods.

Any imposed restrictions for the permissible power levels and the permissible normal operation modes shall be substantiated with references to the relevant sections of the NPP SAR.

16.2.2. Safe operation conditions and configuration of operable systems and equipment required for the NPP power unit start-up and operation in the permissible modes

Information on configuration and state of the systems that should be in operable or readiness condition for the NPP power unit start-up and operation in the permissible modes shall be provided.

Information on the state and configuration of the following systems shall be presented:

- the primary circuit coolant system;

- the secondary circuit heat removal system;

- the third circuit heat removal system;

the in-core instrumentation system;

- the instrumentation and control system;

- the reactor emergency protection system;

- the emergency reactor cooldown system;

- the overpressure protection system of the reactor vessel and its safety containment;

- safety containments of the reactor, the SFAD, auxiliary pipelines of the primary circuit, the primary circuit rooms and the special ventilation systems;

- reliable power supply systems.

The following information shall be provided for each system:

- configuration and quantity of the equipment to be operable for start-up and operation in permissible normal operation modes;

- requirements: for the amount and quality of working media; for activation of the equipment including sequence of actions, the operation logic for the automatics and internal protections; for the characteristics of the systems (power, supply, time, etc.); for the supporting safety systems (power supply, cooling systems, ventilation, etc.);

- the operator's intervention conditions.

The permissible duration of the reactor power operation in case of any failure (shutdown for repair) of safety systems (channels) shall be specified with indication of the operator's actions (to reduce the power, to bring into hot shutdown state, to cool down) if the failure (or repair) is not eliminated within the established time period.

The above-mentioned conditions shall be established for the individual normal operation process systems that prevent the RP bringing into safer state with adherence to the design normal operation procedures (i.e. without usage of safety systems) in case of their complete failure.

The safe RP operation conditions in case of a complete failure of the unit control system (a part of the normal operation control system) shall be determined.

Conditions for periodical integrated testing shall be specified for all safety systems.

Safe operation conditions determined by the SS state and frequency of testing shall be substantiated in accordance with the requirements of the GSR and the NPP RF NSR.

16.2.3. Conditions for maintenance, testing and repair of safety-related systems and metal state control for the RP

The conditions of testing, inspections, maintenance and repair of safety-related systems shall be provided, including:

- the reactor pressure vessel and safety containment;

- the primary and secondary circuit pumps;

- rotating plugs, the CPS column;

- the systems and mechanisms for reloading of fuel assemblies;

- neutron monitoring sensor units;

- the CSS;

- the CPS actuators;

- the reactor control and protection systems;

- the intermediate heat exchanger;

- fresh and spent fuel assembly drums;

- hydraulic drives for the check valves of the primary circuit pumps;

- sodium valves;

- the secondary circuit pipelines;

- the ERCS equipment;

- the system for heat removal to the ultimate heat sink;

- the reliable power supply system;

- the SG, the safety valves of the SG and the superheated steam system, BRU-A, BRU-K;

- the reliable service water supply system;

- the feeding units, the deaerator make-up system;

- lining of the primary circuit rooms;

- the SFP lining;

- the floor lining and lining of any other rooms.

16.3. Operation limits and conditions

16.3.1. Operation limits

16.3.1.1. Limit values of the process parameters


- the limit values of the process parameters corresponding to the boundary values of the normal operation area for each system;

- the limit values of the parameters for all equipment included into the system;

- substantiation of the selected parameter values in the permissible modes, accuracy of their measurement, measurement points, redundancy of measuring channels, permissible time of information loss (reference may be given).

In general, the information on the operation limits of the following basic process parameters and condition characteristics of the NPP power unit systems shall be provided:

- activity of the primary and secondary circuit coolant, the primary circuit protective gas, air (or gas) cooling the biological protection and the reactor cavity, the coolant of the spent fuel assembly drum, the water in the spent fuel assembly pool, releases and discharges of radioactive substances into the environment;

- contamination of the coolant with any foreign impurities;

- the damage level for fuel elements in the nuclear core;

- temperature of the coolant and working media in the main circuits;

- gas pressure in the reactor gas blanket and safety cavities;

- the reactor power deviation limits in the automatic or remote power maintenance mode;

- permissible power levels of the reactor for the permissible normal operation modes;

- the change rates of the reactor power, temperature of the coolant and working media in case of the power unit operation mode changes;

- steam (water) pressure in the SG;

- pressure in the deaerator;

- water level in the deaerator, water reserve in the condensate storage tank;

- content of impurities in the coolant and working media;

- permissible number of different thermal cycles for safety-related components and equipment.

16.3.1.2. Process protections, interlocks and automatic controllers with their actuation setpoints

Values of the process parameters for activation of the basic process protections, interlocks and automatic controllers shall be specified. The adopted values of the process parameters shall be substantiated for the permissible modes.

16.4. Administrative conditions and documentation of the information on monitoring of safe operation limits and conditions

The requirements for the power plant administration and workers with regard to compliance with the established safe operation limits and conditions shall be provided.

The administration shall ensure documentation and storage of the information related to safe operation limits and conditions in accordance with the GSR requirements.

17. REQUIREMENTS FOR THE SECTION "QUALITY ASSURANCE"

17.1. General provisions

17.1.1. The requirements for the information on quality assurance for all works and services affecting the NPP safety are specified in this section; the Applicant shall present this information:

within the NPP PSAR - at the stages of preliminary site approval and obtaining of the NPP construction license;

within the NPP SAR - at the stage of the operation license obtaining.

17.1.2. The information presented by the Applicant within Section 17 of the NPP PSAR or the NPP SAR shall ensure that the design, construction and operation of the NPP under consideration are carried

out (will be carried out) properly and comply with the specified quality assurance requirements.

17.1.3. Information on the areas of activity described in Subsection 17.2 shall be provided at the relevant licensing stage in order to assess acceptability of the Applicant's quality assurance activities.

It is permitted not to provide the information on the areas of activity described in par. 17.2.5 - 17.2.12 in case no works have been performed in these areas.

17.1.4. Section 17 of the NPP PSAR or the NPP SAR (hereinafter - the NPP SAR) shall be divided into subsections with the names corresponding to the areas of quality assurance activities in accordance with Subsection 17.2.

The information provided in Section 17 of the NPP SAR shall be prepared with due regard for analysis of the developed quality assurance programs and their implementation as of the SAR development moment in accordance with the requirements stated in Subsection 17.2.

17.1.5. The regulatory documents used for development and implementation of the quality assurance measures in this area shall be specified for each area of the quality assurance activities presented in the NPP SAR in accordance with Subsection 17.2.

17.1.6. The following shall be provided together with the NPP SAR:

1) at the preliminary site approval stage:

the general quality assurance program for the NPP;

the quality assurance program for the NPP power unit site selection; the quality assurance programs for selection of the sites for NF and RW storage facilities, RW processing plants at the NPP territory;

2) at the construction license obtaining stage:

the quality assurance program for the NPP power unit design;

the quality assurance program for the NPP power unit RP development;

the quality assurance program for the NPP power unit construction;

the quality assurance program for the NPP commissioning;

the quality assurance programs for construction of the NF and RW storage facilities, RW processing plants at the NPP territory;

the quality assurance programs for commissioning of the NF and RW storage facilities, RW processing plants at the NPP territory;

the list of the quality assurance programs developed and planned to be developed for design and manufacturing of the safety-related NPP equipment, items and systems;

3) at the operation license obtaining stage:

the quality assurance program for the NPP operation;

the quality assurance programs for operation of the NF and RW storage facilities, RW processing plants at the NPP territory.

17.2. Information on the quality assurance activity areas

17.2.1. Quality assurance policy

The quality assurance policy adopted by the operating organization shall be described.

It should be demonstrated that the quality assurance policy is coordinated with other areas of the operating organization activities, communicated to all executives and establishes the following:

- principles and objectives adopted in order to ensure safety and foreground in relation to any other objectives;

- the main quality assurance objectives;

- tasks aimed to achieve the established quality assurance objectives and the methods to solve them;

- liabilities of the operating organization management.

Information confirming implementation of the operating organization policy in the area of quality assurance shall be provided.

17.2.2. Organizational activities

17.2.2.1. Quality system of the operating organization


- the quality system structure;

- the list of the main quality system documents (quality guidelines: the general guidelines and the guidelines for individual areas of activities, etc.);

- regulatory, organizational and methodological framework of the quality system;

- responsibility of the parties for quality assurance;

- structure of the quality departments;

- authorities, responsibility, direct functional duties immediately performed by the operating organization;

- infrastructure of the operating organization formed by specialized enterprises and organizations to which it partially delegates its functional duties, authorities and responsibility retaining overall responsibility without any prejudice for liabilities and legal responsibility of contractors;

- distribution of responsibility for quality assurance between organizations performing works and rendering services to the operating organization;

- the procedure for distribution of works affecting the NPP safety assurance and interaction between production units of the operating organization or organizations performing the works and rendering services to the operating organization in the course of these works as well as registration of such interaction in the provisions on production units, job descriptions for the workers and (or) any other organizational and administrative documents;

- the reporting documentation containing efficiency analysis for the quality system of the operating organization, results of its inspections and corrective measures.

17.2.2.2. Arrangement of works

The organizational structures, description of the functional duties demonstrating the authority levels and the internal and external communication lines shall be provided.

The subsection shall contain the following:

- structure of the organizations and departments assuring quality as well as any other functional organizations performing any activities affecting quality in the course of design, manufacturing, construction, installation, commissioning works, testing, inspections and revisions of the reporting documentation;

- the general design arrangement scheme demonstrating interaction of the operating organization, the main organization for the NPP design development and their contractors as well as the design approval procedure;

- the list of documents defining the form of incorporation for the operating organization and the NPP;

- the procedure for the NPP SAR development and issuance at different licensing stages;

- information on the control system of the operating organization and the communication lines between the operating organization and its contractors for all quality assurance works as well as their efficiency for implementation of the general and individual NPP QAPs;

- the list of managerial positions with the authorities and responsibility for implementation and efficiency of the general and individual NPP QAPs.

17.2.2.3. Quality assurance programs


- information on development, issuance and inspection results with regard to implementation of the general and individual quality assurance programs in accordance with the established requirements for the NPP quality assurance program;

- information on implementation of the general and individual NPP QAPs as of the moment of the NPP SAR submittal by the Applicant;

- information on compliance of the NPP QAPs with the established requirements for the NPP quality assurance program;

- the NPP QAP application scope;

- information confirming that any activity affecting safety-related systems and equipment is subject to the relevant control within the NPP QAP framework;

- description of the measures implemented prior to the NPP SAR submittal for review (particularly terms of reference for feasibility studies, development of the RP, the NPP construction project, etc.);

- description of the measures taken by the operating organization in order to ensure current implementation of the NPP QAP;

- information on analysis of the regulatory and technical support performed by the operating

organization at all NPP construction and operation stages;

- description of the measures taken by the operating organization in order to ensure development of the missing regulatory documents as detected in the course of analysis.

17.2.3. Staffing and training of the workers

The subsection shall contain information on the procedures adopted in the operating organization to work with the employees with regard to:

training, knowledge and skill checks for the workers engaged in any works affecting the NPP safety assurance and supervision over these works (particularly the workers carrying out tests, inspections and checks);

determination of the needs for training of the workers and organization of training, retraining, development of competence and certification of the workers including issuance of the relevant certificates;

analysis of programs for training, retraining, development of competence and certification of the workers;

maintaining of records on training, retraining, development of competence, and certification of the workers.

17.2.4. Regulatory documents

The subsection shall contain the list of regulatory documents on quality assurance effective in the operating organization (or references thereof). for example federal regulations and rules in the area of atomic energy use, national and industrial standards, corporate standards, effective procedures of the quality system.

The quality system procedures aimed to ensure compliance with the requirements of the report and the adopted quality assurance policy shall be specified.

17.2.5. Documentation management

The subsection shall contain the information on the procedures adopted in the operating organization for development, agreement, approval, implementation, identification, introduction of changes, review, distribution, storage, disposal of ineffective documents (drawings, manuals, methodologies, data. etc.).

17.2.6. Design control


- description of the measures (procedures) planned and implemented by the operating organization within the framework of the general quality assurance program for design control that shall prescribe correctness verification for the applied solutions as well as their compliance with the design requirements;

- analysis of feasibility and subsequent implementation of the initial design requirements within the terms of reference for the NPP design, development of the RP and equipment (in this case attention shall be paid to the requirements for safety and reliability);

- description of the design verification methods applied by the operating organization;

- verification of compliance with the requirements for documenting of the inspection results so that to enable investigation or revision of the verification method after its completion;

- verification of compliance with the requirements for time limits of the inspections that shall be completed after prototype or commercial prototype testing prior to issuance of any documentation for manufacturing or construction;

- check of compliance with the criteria of mandatory testing provided for verification of the design, the necessity to ensure representativeness of tests and modelling of the least favorable conditions defined on the basis of safety analysis;

- description of the measures aimed to define and control delimitation of works in the course of design;

- information on availability and implementation of the procedure for control of any change introduction to the design in the course of design and manufacturing at the NPP construction site as well as in the course of the NPP operation.

17.2.7. Procurement management for equipment, components and materials as well as provided services

The subsection shall contain information on the following procedures effective in the operating organization:

- arrangement of procurement of equipment, components and materials, as well as provision of services including the procedure for selection of organizations performing works and rendering services to the operating organization;

- maintenance of the procurement documents for equipment, components and materials as well as for provision of services;

- inspection of the quality assurance programs of the organizations performing any works and rendering any services to the operating organization, assessment of their capability to perform the works or render the services to the operating organization;

- analysis of the contracts for procurement of equipment, components and materials as well as for provision of services.

17.2.8. Control of the purchased equipment, components and materials and provided services

The subsection shall contain information on the following effective procedures:

- arrangement of identification, control (particularly incoming one) and testing of the equipment, components and materials;

- assurance of traceability for the control and testing results;

- assurance of completeness for various types of control and testing;

- arrangement of storage, transportation, preservation and packing of the equipment;

- arrangement of supervision over compliance with the requirements for the provided services.

17.2.9. Industrial activities of the operating organization and any organizations performing works and rendering services to the operating organization

The subsection shall contain information on the effective procedures for performance of the required operations aimed to control quality of the processes important for the NPP safety assurance, particularly the following processes:

- mechanical processing and assembly of the equipment and units of safety-related systems;

- cleanness assurance in the course of manufacturing;

- installation and dismantling of the equipment, units, civil structures affecting safety;

- non-destructive control methods.

- welding, surfacing, heat treatment;

- repair of the equipment and maintenance in the course of operation.

The following information shall be also presented in the subsection:

- development of the list of safety-related systems (components);

- presence of any requirements for quality of safety-related systems (components) and works affecting the NPP safety assurance;

- the procedure and techniques for performance and control of the works affecting the NPP safety assurance;

- application of the statistical methods (in case of necessity).

17.2.10. Inspection control

The subsection shall contain information on the results of implementation of the general and individual quality assurance programs by carrying out inspections, particularly:

- the lists of inspections;

- the inspection programs;

- the inspection planning schedule and its implementation;

- confirmation of independence of the inspectors from the inspected works;

- availability and implementation of the NPP QAP;

- the instructions on the procedure for inspections of the process control points, stages of the works after which any further works are prohibited without inspection and documented permit based on the control and inspection results.

17.2.11. Control of testing

The subsection shall contain information on the effective procedures ensuring the complete set of test types and trial of the equipment, items and systems important for the NPP safety.

The following shall be also provided:

- the list of tests for the equipment and systems in order to check their operability in the course of operation;

- information on the ways to reflect the product operation model, requirements for metrological support, acceptability conditions for the testing results and representativeness of tests in the testing programs;

- methods for registration and documentation of the testing results and assessment of their acceptability;

- references to the testing reports and description of the testing results with due regard for implementation of the general and individual quality assurance programs.

17.2.12. Metrological support

The subsection shall contain the following information:

- development and implementation of the verification program for the measuring equipment and instruments as of the moment of the NPP SAR submittal;

- the application scope of the verification program, availability of the lists of inspected equipment and instruments;

- availability of the provision for identification of the measuring equipment and instruments;

- the effective procedures for:

arrangement of validation, calibration, verification and identification of the measuring and testing equipment and instruments;

maintenance of good operating condition and servicing of the measuring and testing equipment and instruments;

maintenance, accounting and storage of the validation, calibration and verification reports for the measuring and testing equipment and instruments.

17.2.13. Quality assurance for software and calculation methods


- information on the effective quality assurance programs for software and calculation methods, particularly verification of software and calculation methods;

- the list of programs used for engineering calculations (physics, thermal hydraulics, strength, etc.), design and research works with indication of the program capabilities, results of their verification, information on registration and validation (certification);

- the procedure for arrangement and quality assurance of the calculation works;

- the procedure for updating of the technology for calculation substantiation of structures at all design stages;

- information on the competence development for the work performers;

- information on usage of any validated databases for development of programs;

- information on assimilation and implementation of alternative domestic and foreign programs;

- the procedure for training of the work performers in state-of-the-art methods for numerical solution of thermophysical and other problems;

- the procedure for validation of software.

17.2.14. Reliability assurance

The subsection shall contain information on the effective procedures aimed to ensure reliability of the safety-related NPP equipment, items and systems as well as the procedure of interaction and the organizational chart for the participants of the reliability assurance works.

17.2.15. Control of non-conformities

The subsection shall contain information on the following effective procedures:

- registration of any deviations from the requirements for quality of works (services) and (or) equipment (design or manufacturing errors, defects and failures of the equipment, abnormal operation modes, human errors, etc.) and their analysis;

- prevention of usage of any products that do not comply with the established requirements (for example, the procedure for segregation, disposal, documenting, identification of such products) or acceptance of any services that do not comply with the established requirements;

- arrangement of the system for collection and processing of data on any detected non-conformities, violations, defects, their causes and the implemented corrective measures;

- detection, registration and informing of the relevant organizations on any revealed deviations in the materials, equipment and components.

Information on any registered cases of decision making on the detected non-conformities, results of their control by the quality departments and analysis of the detected deviations by the operating organization shall be also provided.

17.2.16. Corrective measures


- information on the effective procedures for development of corrective measures aimed to prevent repeated non-conformities, particularly subsequent to the inspection results, control of their implementation, assessment of their efficiency and documenting of these activities;

- information on the effective procedures for prevention of any potential deviations and non-conformities and control of their efficiency assurance;

- the list of the basic corrective measures subsequent to implementation of the general and individual quality assurance programs as of the moment of the NPP SAR submittal.

17.2.17. Quality program records


- the procedure for control of information on quality assurance in the operating organization and at the NPP in relation to the NPP quality assurance issues;

- availability and implementation of the procedure for recording, storage and issuance of documentation that shall be maintained in accordance with the written procedures;

- information on the effective procedures for compilation and maintenance of the quality assurance documentation (establishment of the record types depending on importance, identification, collection, indexing, access, filing, storage, maintenance and destruction of the registered quality data including the results of inspections, testing, verification of the processes, analysis of the supplied equipment, components and materials);

- description of the NPP QAP implementation reporting system which shall include inter alia the procedure for compilation of:

reports on the results of any inspections carried out with regard to application of documents, quality of the developed products, quality costs, credibility assessment, etc.;

annual reports on the product quality for the certain period;

annual reports on the results of the designer's supervision in the course of manufacturing, installation, testing and operation.

17.2.18. Inspections (audits)

The subsection shall contain information on the effective procedures for performance and presentation of the results of any independent inspections (internal and external) of the actual quality assurance program implementation state as well as its efficiency assessment.

18. REQUIREMENTS FOR THE SECTION "DECOMMISSIONING OF THE NPP POWER UNIT"

18.1. Decommissioning concept

The strategy of the NPP power unit decommissioning shall be considered by review of various NPP power unit decommissioning options with description of the potential final states of the power unit for each option.

The concept shall demonstrate the way to provide the following at all implementation stages of the NPP power unit decommissioning options:

- reduction of dose burdens for the personnel and the public in accordance with ALARA principle;

- generation of the minimum RW amounts (volumes);

- reduction of RSb ingress into the environment down to the minimum achievalbe level.

The concept shall demonstrate that the following is provided in the power unit design:

- selection of the structural and construction materials with the minimum activation by neutrons in the course of further power unit operation;

- the design solutions limiting the possibility for transfer and propagation of activated corrosion products in the process media (coolant, moderator);

- location of the rooms in the buildings and facilities enabling to carry out decontamination of the surfaces (external and internal) in the best way for all decontamination methods;

- reservation of the areas at the power unit site or at the NPP site for storage of radioacyive wastes and reusable materials generated in the course of the NPP power unit decommissioning;

- the possibility to locate special-purpose equipment required for the NPP power unit decommissioning in the power unit rooms.

Selection of the most feasible NPP power unit decommissioning option shall be substantiated with due regard for the following main factors:

- the expected engineering and radiological state of the power unit as of the final shutdown and the possibility to predict its state at the required moment within the entire decommissioning period;

- assessment of the possible hazardous radiation impact on the workers, the public and the environment;

- requirements of the safety RD effective as of the moment when the requirements for the report are developed;

- assessment of the amounts, types, physical state of the radioactive wastes generated in the course of decommissioning;

- availability of the plants and techniques for handling of RW generated in the course of the NPP power unit decommissioning;

- availability of storage facilities for storage and disposal of RW generated in the course of the NPP power unit decommissioning;

- assessment of the amounts, types and physical state of the materials;

- potential prospects and plans for further usage of the decommissioned power unit site;

- financial capacities of the operating organization for the NPP power unit decommissioning;

- potential prospects for economic and social development of the region.

18.2. Design solutions

The technical solutions adopted in the design and aimed to ensure safe performance of the future NPP power unit decommissioning works shall be analyzed, including the following information:

- the load-bearing capacity of the civil structures, buildings and facilities not only within the design service life but also for the NPP power unit decommissioning period;

- operability of the systems required not only within the design service life of the power unit but also for the NPP power unit decommissioning period or the possibility of their replacement or life extension;

- the plants and techniques for handling of RW generated in the course of the NPP power unit decommissioning;

- the transportation scheme for delivery of RW from the main building to the RW handling complex and the storage facility for conditioned RW;

- transportation of the dismantled radioactive equipment to the RW handling complex for decontamination, fragmentation and conditioning with subsequent safe relocation for disposal in the regional or central repositories;

- the reserve of additional areas for location of the facilities required at the NPP power unit decommissioning stage provided on the general layout of the NPP site;

- removal of the wastes accumulated in the course of the power unit operation from the RW storage facilities prior to commencement of the NPP power unit decommissioning works;

- reservation of the areas for storage of reusable materials (for restricted and unrestricted use) at the NPP site;

- arrangement of the outgoing radiological control of the materials returned for reuse in the national economy.

The design requirements for the civil structures aimed to facilitate their dismantling in the future shall be analyzed. These requirements shall include the following:

- arrangement of the civil structure fragments with the geometric dimensions enabling to divide its activated part according to the induced activity levels (high, medium, low) as well as into the parts for restricted and unrestricted use;

- application of the modular civil structures and process equipment (for example, the primary circuit of the reactor) for radiation protection while ensuring all strength characteristics of the protective structures;

- arrangement of the modular option for the protective structures providing for the possibility to divide them into the zones with and without RSb contamination;

- use of special-purpose single- and multi-layer sealing coatings in order to reduce radionuclide contamination of the concrete structures, selection of the concrete components in order to reduce the depth of radionuclide penetration into the concrete;

- application of side-out panels in the floors and walls of the rooms in order to arrange installation openings facilitating access to radioactive equipment.

Sufficient capacity of the regular power unit ventilation, water supply, air purification systems, etc. for performance of the complete scope of dismantling works in the future shall be analyzed.

Arrangements for radiation safety assurance in the course of the NPP site rehabilitation shall be analyzed.

18.3. Characteristics of the equipment, buildings, facilities, civil structures

The mass and volumetric characteristics of the main process and auxiliary equipment of the power unit shall be specified and analyzed:

- dimensions, weight and material of the fast reactor pressure vessel and safety containment, the total weight of the fast reactor without sodium, the sodium volume in the primary and secondary circuit, the reactor nuclear core dimensions, the number of fuel assemblies in the nuclear core, the number of fuel assemblies in the breeding blanket, the reactor cavity dimensions and weight, etc.;

- dimensions and weight of the primary circuit RCP, the RCP electric motor, the large rotating plug, the relay protection, the "sodium-sodium" heat exchanger, intermediate "sodium-sodium" heat

exchangers, the SG, the emergency primary circuit sodium discharge tanks, the feedwater plant, low-pressure heaters, high-pressure heaters, etc.;

- dimensions and weight of the primary, secondary and third circuit equipment and pipelines;

- dimensions and weight of the equipment of the refueling, sodium purification, sodium fire extinguishing systems, etc.

Chemical composition of the materials required to calculate neutron activation of the equipment assemblies and conponents, engineering and civil structures shall be provided.

The range of the process and auxiliary power unit equipment, its configuration, mass and volumetric characteristics, grades and chemical composition of steels it is made of shall be provided.

Information on the suitability (for example, fragmentation, etc.) and limitations of the main, process and auxiliary equipment for dismantling and transportation in the course of the NPP power unit decommissioning and on the range and characteristics of additional hoisting and transportation means and utilities required for performance of the NPP power unit decommissioning works shall be provided.

This information shall:

- correlate to the adopted design and engineering solutions aimed to ensure safe decommissioning of the NPP power unit and inform about the planned basic solutions aimed to ensure safe decommissioning of the NPP power unit and related to the NPP general layout, the NPP site, the power unit site, transportation of the dismantled equipment, systems and civil structures (including radioactive ones), RW processing and storage at the NPP site;

- inform on the spatial and planning solutions adopted in the design of the power unit in order to ensure performance of dismantling operations and transportation through the use of remotely controlled means including robotic ones;

- inform on the selected structural materials for operation under the radiation conditions and aimed to reduce formation of long-lived radionuclides;

- inform on the use of detachable and modular assemblies in the main equipment and systems of the power unit, detachable joints and connections of the equipment parts with different RSb contamination degree.

The following information shall be also presented in the subsection:

- application of easily removable coatings and other means in the power unit design and implementation of the measures for limitation of radioactive contamination propagation and fixation;

- provision of the possibility to take samples of the load-bearing metalworks in order to determine the actual mechanical properties;

- deep decontamination of the non-reusable equipment and systems and provision of the required areas (rooms) for temporary stocking and orderly storage of radioactive wastes generated in the course of the NPP power unit decommissioning.

18.4. Assessment of qualitative and quantitative composition of radioactive substances accumulated in the power unit equipment and civil structures

Conservation calculations of the radionuclide content in the materials of the equipment and civil structures due to their neutron activation (for example, at the moment of final shutdown or in a year

after shutdown) shall be provided based on the data (the range of equipment and structures, mass and volumetric characteristics, chemical composition of the materials, etc.) specified in the previous subsections.

The total and volumetric specific activity data shall be presented for each power unit equipment and structure item subjected to neutron irradiation in the course of the power unit operation. Information on the distribution of radionuclides (activation products) with regard to depth shall be provided for protective civil structures (for example, the reactor cavity, biological protection, etc.). Conservative assessment of RW and reusable materials (with regard to weight and volume) shall be provided based on this information.

The subsection shall include the analysis results for usage of two possible design options to reduce the quantity of radionuclides in the steel structures of the power unit with the BN reactor:

- replacement of alloys containing great amounts of cobalt and nickel with the alloys low in these components or alloys without these components;

- decrease of cobalt, silver, niobium and nickel content in the structural materials.

The issue on limitation or complete elimination of serpentinites, chromites, magnetites in the protective civil structures due to high content of cobalt and iron in them shall be analyzed, or their application shall be substantiated.

The issue on lithium content in the materials of civil structures shall be considered, as lithium becomes the source of tritium generation due to neutron absorption. All significant paths of tritium generation shall be analyzed.

The quantity of radionuclides and particle size of aerosols generated in the course of dismantling works shall be assessed on the basis of proposed equipment fragmentation and concrete structure destruction technologies.

The gamma radiation dose rates due to individual activated assemblies of equipment and structures shall be assessed for the industrial rooms and the power plant site as of the same moment as for the activity assessment.

18.5. Radiological control in the course of decommissioning

The requirements for the scope of radiometric, spectrometric and dosimetric control shall be defined, and the analysis in comparison with the radiological control solutions provided in the design for the power unit operation stage shall be performed based on the analysis of ionizing radiation sources and characteristics of aerosols. It should be demonstrated that the radiological control system provided in the design complies with the requirements listed below and can be used after the power unit shutdown within the entire NPP power unit decommissioning period.

For this purpose it shall be demonstrated that the radiological control system can ensure the following measurements:

- specific activity of any radionuclide within the range of 10 Bq/g to hundreds kBq/g for segregation of the materials into wastes (low-, medium- and high-level) and reusable materials (for restricted and unrestricted use);

- gamma radiation dose rates in the rooms within the range of 0 to 100 R/h;

- gamma radiation dose rates for individual assemblies and equipment of the reactor internals, the reactor vessel, etc., fragments of the fast reactor equipment in the course of dismantling, segregation in

accordance with the radioactivity groups and transportation;

- surface beta contamination of the equipment and rooms;

- specific volumetric activity of aerosols;

- specific volumetric activity of aerosols in the vent stack.

The range of measured energy values for gamma quanta (photons) shall be within 0.015 to 3 MeV.

It should be demonstrated that the external dosimetry system ensures monitoring of ingress to the environment for any radionuclide generated in the course of the NPP power unit decommissioning works or any mixtures thereof.

In case these requirements are not met it should be stated that these issues shall be solved in the NPP power unit decommissioning project.

18.6. Informational support of decommissioning

The Applicant shall provide information confirming that it will perform safety analysis for the NPP power unit decommissioning to be used as the basis to apply for the license in order to perform these works.

Appendix 1

STANDARD STRUCTURE FOR DESCRIPTION OF SYSTEMS IN THE NPP SAR

1. Purpose and design basis

The purpose of the systems shall be specified, safety classification, safety classes in accordance with the GSR and groups in accordance with the NPU Rules shall be indicated for the components, etc.

The list of regulatory documents on safety containing the requirements that the described system shall comply with shall be provided, the principles and criteria used as the basis of the system design shall be specified.

The design basis for the systems shall be presented.

The material shall be presented in the following sequence:

- purpose and functions of the system;

- design basis.

2. System design

Description of the design and (or) process flow diagram of the system in general, its sub-systems and components (if they perform independent functions) shall be presented. Drawings, figures and schemes illustrating the design and operation of the system, its spatial layout and interfaces with other NPP systems in sufficient detail shall be provided.

The basic technical characteristics of the system and its components shall be specified.

Selection of the materials shall be substantiated with due regard for normal operation conditions and operational occurrences including emergency situations and accidents, and information on validation of the materials shall be provided.


- description of the process flow diagram;

- description of the equipment and components;

- description of the used materials;

- overpressure protection (if needed);

- layout of the equipment.


The list and substantiation of the permissible values for the controlled parameters of the system in all operation modes shall be provided; location of the measuring points shall be specified, the control methods shall be described, and information on metrological validation of the applied methods and requirements for the instrumentation shall be presented. The system interfaces with control systems, redundancy of sensors and communication channels shall be described. Complete lists of the measuring points and sensors, protections, principal description of the controllers, diagnostic systems and automatic control programs shall be provided.


- the measuring points, the list of controlled items with indication of the control method (the operator, automatics, the equipment protection method, the safety protection method);

- description of protections for the controllers, diagnostic systems, automatic control programs.


The controlled parameters in the course of manufacturing, construction and installation of the NPP systems and components shall be provided.

The list of nuclear-hazardous works in the course of installation, testing, operation and repair shall be presented.

The scope and methods of incoming control, commissioning tests, tests and inspections in the course of operation and their metrological support shall be substantiated; the list and permissible values of the controlled parameters and the requirements for the instrumentation used for the purpose of testing shall be presented and substantiated.

5. System analysis

Description and algorithms of the calculation programs used to analyze operation of the system, the input data for the calculations, assumptions and restrictions, the calculation results and conclusions shall be provided. Information on verification and validation of the calculation programs shall be presented. The scope of information shall be sufficient to perform independent alternative calculations in case of necessity. In case any experiments were performed in order to substantiate safety of the system the conditions of the experiments shall be specified, their compliance with the design conditions shall be analyzed, the experimental facilities and metrological support of the experiments shall be described, and interpretation of the results with regard to the design conditions shall be provided. Information on certification of the equipment, systems and components shall be provided.

Description of the system functioning under normal operation conditions and in case of any operational occurrences including near-accident situations and design basis accidents, interactions with other systems with due regard for their possible failures and measures for protection of the system against the effects of these failures shall be provided. Operation limits and conditions, safe operation limits and conditions, SS actuation setpoints and reliability parameters of the system and its components shall be specified for the provided operation modes.

Failures of the system components including human errors and impacts of failure consequences (particularly common cause failures) on operability of the system under consideration, any interfaced systems and the NPP safety in general shall be analyzed. Failures requiring special consideration in Section 15 shall be specified.

Compliance of the system design with the RD requirements shall be analyzed.


- normal operation including the data on calculation programs;

- safe operation limits and conditions;

- functioning of the system in case of failures;

- functioning of the system in case of emergency situations and design basis accidents;

- functioning of the system under external impacts;

- reliability parameters of the system;

- certification of the equipment, systems and components.

In the course of information presentation references to other sections where this information is provided in more details are possible.

Particular content of each subsection may change depending on the system peculiarities.

Individual subsections may be omitted or supplemented with other ones if it is determined by the system peculiarities.

6. Conclusions

Conclusions on compliance of the system with the requirements of federal regulations and rules in the area of atomic energy use and compliance with the requirements of any other regulatory documents on safety shall be made.

Appendix 2

NUCLEAR POWER PLANT SITING CONDITIONS

1. General information1.1. NPP name/power unit number ___________________________/_____1.2. Power unit commissioning/ decommissioning year ___/____1.3. Location Constituent entity of the Russian Federation

________________________________ The nearest city (cities) _______________________________________ Distance from the site to ____________ km Azimuth angle (degree) __________________1.4. Geographical coordinates of the site (the NPP center) Latitude ______________________________ Longitude _____________________________1.5. Absolute elevations of the site in the Baltic height system (BS) Natural: highest/ medium/ lowest ___/____/___ m BS Levelling _________________ m BS1.6. Landscape within the radius of 20-30 km Brief description Plain _____________________________________________________ Hilly terrain _________________________________________ Position in a valley __________________________________________ Location of rivers ____________________________________________ Lake/ sea coastline __________________________________ Other (please specify) ____________________________________________1.7. Population distribution The nearest administrative center, village, city: Name ____________________________________________________ Distance/azimuth _________ km/_____ Population ______________ people The nearest large city (> 50000 people) Name ____________________________________________________ Distance/azimuth _________ km/deg. Population ______________ people2. Meteorological conditions 2.1. Tornado hazard area according to the zoning map ______________________ 2.2. Tornado intensity class according to Fujita scale _____________________________________ 2.3. Maximum tornado wall rotational movement velocity _________ m/s 2.4. Tornado path length/ width ________ km/ ________ m 2.5. Pressure differential between the tornado funnel periphery and center _______ hPa 2.6. Probability of a tornado within the NPP site ___________ 2.7. Probability of a hurricane (tropical cyclone) within the NPP site ___________________ 2.8. Estimated characteristics of the maximum hurricane (tropical cyclone) ________________________________ 2.9. Estimated maximum wind speeds with various probability, including 1, 0.1 and 0.01% ___, ___, ___ m/s3. Hydrological conditions 3.1. Type of water body affecting the NPP safety (river, lake, water reservoir, sea offshore zone) _______________________________________ 3.2. MSW formation factors provided in the design For rivers: spring flood, rainfall flood, dam and dyke breakage, blockage, ice jams and gorges, volcanic activities, earthquake, rock fall, landslide, mudflow, etc. (underline as appropriate, specify other factors) ___________________________________________ For water bodies: wind surge, storm waves, maximum wave setup, seiches, tsunami waves, tides, etc. (underline as appropriate, specify other factors) ___________________________________________ 3.3. Absolute elevation of the highest observed (historical) water level in the water body ____________________ m BS 3.4. MSW parameters Maximum levels with various probability, including 1, 0.1 and 0.01% _______, _________, _______ m BS Maximum wave height with various probability, including 1, 0.1 and 0.01% _______, _________, _______ m For rivers: Maximum water flow with various probability, including 1, 0.1 and 0.01%

_______, _________, _______ m3 For water bodies: Absolute MSW level elevation ____________________________ m BS Highest water level in case of a seiche ___________________________ m Maximum amplitude of flood-and-ebb sea variations ____________m Estimated storm setup height at maximum wind speeds with various probability, including 1, 0.1 and 0.01% _______, _________, _______ m Maximum deep-water wave height at maximum wind speeds with various probability, including 1, 0.1 and 0.01% _______, _________, _______ m Highest elevation of the coast flooding with tsunami waves of various probability, including 1, 0.1 and 0.01% _______, _________, _______ m BS Lowest elevation of the coastline dewatering in case of tsunami waves with various probability, including 1, 0.1 and 0.01% _______, _________, _______ m BS4. Hydrogeological and geotechnical conditions 4.1. The first aquifer from the surface Non-artesian/ artesian (underline as appropriate) Distribution area _________________________________________ Absolute elevation of the lower/ upper aquiclude _______ m/ _______ m BS Maximum/ medium/ minimum absolute elevation of the groundwater level _______ m/_______ m/_______ m BS Lithological characteristics of the adjacent rocks _______________________ Filtration coefficient of the rocks ____________ m/day Active porosity of the rocks ____________ % Existing water intake _____________________________________ Maximum/ medium/ minimum absolute elevation of the groundwater level in the RB area _______ m/_______ m/_______ m BS 4.2. The second aquifer from the surface Distribution area _________________________________________ Absolute elevation of the lower/ upper aquiclude _______ m/ _______ m BS Maximum/ medium/ minimum absolute elevation of the groundwater level _______ m/_______ m/_______ m BS Lithological characteristics of the adjacent rocks _______________________ Filtration coefficient of the rocks ____________ m/day Active porosity of the rocks ____________ % Existing water intake _____________________________________ Maximum/ medium/ minimum absolute elevation of the groundwater level in the RB area _______ m/_______ m/_______ m BS 4.3. Aquiclude Distribution area _________________________________________ Absolute elevation of the aquiclude top/ bottom __ m/__ m BS Lithological characteristics of the aquiclude rocks ______________________________ Filtration coefficient of the rocks ____________ m/day Presence of any hydrogeological windows in the aquiclude ________________________ 4.4. Geotechnical conditions Specific soils: weak with the deformation module of < 20 MPa, liquefiable, subsidental, swelling, saline, permafrost (underline as appropriate, specify any other soils) __________________________________________________ Hazardous present-day geological processes and phenomena: landslides, karst, suffosion, karst and suffosion, etc. (underline as appropriate, specify any other conditions) __________________________________________5. Seismic activity 5.1. Geodynamic model of the region and the NPP site location area 5.2. Seismotectonic model of the region and the NPP site location area 5.3. Seismological model of the region and the NPP site location area 5.4. Detailed seismic zoning scheme of the region 5.5. Scheme of structural and tectonic conditions in the area 5.6. Seismic micro-zoning scheme of the site for natural and anthropogenically modified conditions 5.7. Spectral structure characteristics and duration of vibrations for distant, intermediate and local earthquakes 5.8. Parameters of SSE and DBE from the nearest seismogenic zones:

magnitude, focal depth, distance to the seismogenic zone r, seismicity J according to MKS-64 scale for benchmark soil of the site. The information shall be presented in the form of a table. The approximate format of the table is given below.

Seismogenic zone number

Magnitude Focal depth, km r, km J, points

SSE DBE SSE DBE SSE DBE SSE DBE

5.9. Seismicity of the RB area in case of SSE/ DBE ______/________ points 5.10. Maximum amplitude of horizontal vibrations on the free-form surface of the RB area levelling in case of SSE/ DBE: accelerations _____/_____ m/s2; velocities ______/______ cm/s 5.11. Maximum amplitude of horizontal vibrations in the top rock formations in case of SSE/ DBE: accelerations _____/_____ m/s2; velocities ______/______ cm/s 5.12. Maximum acceleration/ velocity amplitude periods at the levelling level in case of SSE ______/______ s 5.13. Ratio between vertical and horizontal acceleration ______6. Aircraft crash 6.1. Minimal distance from the site to flight courses and approach routes of any airport ____, _____, ______ km 6.2. Distance to a large airport _________ km 6.3. Probability of aircraft crash at the site _________________________________ The information shall be presented in the form of a table. The approximate format of the table is given below.

Crash category

Probability of aircraft crash at the site, 1/year

prediction in 10 years prediction in 50 years

7. Emergency explosions outside the site within the radius of 10-20 km 7.1. PADS Components of chemical and oil refinery plants; storage facilities for energy products and explosive substances; pipelines for liquid and gaseous energy products (surface-mounted); military facilities (underline as appropriate) 7.2. Surface transport PADS Movement routes, ports, harbours, channels, railway stations, characteristics of traffic flows Appendix. General plan (scale 1:25000)

8. Fires outside the site (within the radius of 2 km) Potential fire sources: forest, peat bog, gas, oil or product pipeline, base/ warehouse/ storage facility for flammable materials, navigable canal (underline as appropriate) Appendix. Topographic and landscape map of the area with indication of the potential stationary fire sources

9. Toxic releases into the atmosphere Sources of toxic vapor/ gas/ aerosol, precipitation releases outside the site (underline as appropriate) Appendix. Layout plan of the release sources

10. Information on natural radioactivity of the NPP location area

Appendix 3

ANALYSIS RESULTS FOR SCENARIOS OF NATURAL AND HUMAN-INDUCED INITIATING EVENTS

No. Initiating event

Initial impacts Secondary impacts

The list of buildings, structures, systems and

components that can be affected

by impact

Need for stability analysis

1 2 3 4 5 6

1. External impacts

1.1. Earthquake of any genesis

Vibration of the base, deformation of the base

1. Vibration of buildings and structures2. Missiles3. Vibration of systems and components

All NPP systems and components

Yes

1.2. Etc.

2. Internal impacts caused by emergency situations at the site

2.1. Explosion of receivers for hydrogen and other explosive gases

4. ASW5. Missiles6. Fire

1. Damage of buildings and structures2. Missiles

Individual systems and components

Yes

2.2. Etc.

3. Internal impacts caused by emergency situations within the site external in relation to the reactor building

3.1. Fire in the turbine hall

Fire load 1. ASW2. Missiles

1. Containment2. Pipelines

Yes

3.2. Etc.

4 Internal impacts caused by emergency situations inside the reactor building

4.1. Pipeline rupture

1. Missiles2. Jets

1. Damage of structures2. Missiles

1. Equipment2. Reactor, etc.

Yes

4.2. Etc.

Note. In case safety-related systems are specified in Column 5 "Yes" shall be written in Column 6. In accordance with the mark in Column 6 results of quantitative assessment of the event probability, parameters of impacts on the affected systems and components and conclusions on stability of these systems and components under these impacts shall be provided in the relevant sections and chapters of the report.

Appendix 4

REQUIREMENTS FOR INFORMATIONON RADIOLOGICAL CONTROL PROGRAMS

The radiological control programs shall contain the following information for each section of the program (references to the information presented in par. 11.6.2 may be given in the programs):

1. Objects of control

2. Control means including their metrological support In this case the following shall be specified:

- types of stationary, portable and laboratory equipment and instruments used for dosimetric and radiometric monitoring, control of surface contamination levels, content of volatile and gaseous radioactive substances in the atmosphere of the rooms; sampling, individual monitoring of the personnel under normal operation conditions, in the course of repair and in case of accidents;

- information on the ways to provide the possibility for redundancy (with regard to the number and locations in case of accidents) of the measuring channels, means for presentation and documentation of information on the radiation situation in the rooms and at the NPP site with display of information in the emergency response control center within the sanitary-protective area.

3. Software and mathematical support Special attention shall be paid to the possibilities to predict the radiation situation and propagation of radioactive substances in the NPP power unit rooms, at the site and in the environment on the basis of advanced mathematical and physical modeling methods under normal conditions, as well as to predict the radiation situation within the entire radiation accident zone in accordance with the list of beyond design basis accidents considered in the design. It shall be demonstrated how geographical conditions, meteorology and development of the adjacent territories is taken into account in the calculations.

It should be demonstrated how the prognostic mathematical models are implemented through the use of application programs on the computers applied in the radiological control system.

4. Computer equipment and methods for processing, analysis, presentation and transmission of information Capabilities of the computer or computer networks used in the radiological control system shall be described. It shall be demonstrated that they are sufficient for prediction of RSb propagation and the radiation situation dynamics within the entire radiation accident zone for the minimum time required to solve this task.

5. Scope and frequency of control for radiological, chemical and meteorological parameters

Appendix 5

INPUT DATA FOR CHECK CALCULATIONS

1. Geometrical input data

The basic structural characteristics (volume, length, flow passage area, elevation differences, heat exchange surface, weight, wall thickness, hydraulic diameters, local resistance, etc.) shall be specified for the equipment performing its functions in case of the corresponding emergency situations and accidents.

2. Physical input data


1. Neutron and physical characteristics (variation and reactivity coefficients, integral efficiency of the CPS, prompt neutron lifetime, share of delayed neutrons, etc.)

2. Thermal and physical characteristics (thermal conductivity, heat capacity and density of the applied materials; temperature and enthalpy of different make-up sources and storage tanks; position of phase levels and mass in vessels with phase separation)

3. Physical and chemical properties of reagents and solutions generated in the course of accident, their radiation stability, partition constants and chemical reactions with basic iodine compounds

3. Process input data

The design characteristics (operation algorithms, setpoints, characteristic parameters, characteristics of the main equipment - pumps, modular devices, etc.) shall be provided.

4. Initial conditions

The presented list of initial conditions for the analyzed process shall be conservative. Conservatism of the initial conditions shall be substantiated.