slide presentation for july 24, 2008 meeting on nuscale ... · severe accident mitigation and...

PRE-APPLICATION MEETING

July 24, 2008

U.S. Nuclear Regulatory CommissionPre-Application Meeting

Rockville, MD

Company Overview and Product History

Paul LorenziniCEO

July 24, 2008


Rockville, MD

ObjectivesIntroduction to NuScale Power

The Product

The Company

Strategic Relationships

The Market

Approach to Pre-Application Reviews

3

●Construction Simplicity:● Entire NSSS is 60’ x 15’.

Prefabricated and shipped by rail, truck or barge

●Natural Circulation cooling: ● Enhances safety –

eliminates large break LOCA; strengthens passive safety

● Improves economics –eliminates pumps, pipes, auxiliary equipment

The Product: NuScale Power is Innovative…

4

…Yet relies on proven LWR Technology

Light water technology utilizes large existing base of R&D

NuScale can be licensed within existing regulatory framework

Fully integrated prototype test facility available for licensing

“Off-the-shelf” systems (turbine-generators; fuel) facilitate commercialization

5

The NuScale Product: HistoryThe NuScale concept was originally developed through a DOE funded effort between 2000 & 2003

Called MASLWR (Multi-Application Light Water Reactor), it was targeted to serve markets in developing countries

Results published in 2003

Significant proprietary improvements by Oregon State University since 2003

Patents filed in November 2007

6

NuScale Power Inc: HistoryCompany formed – June 2007

Initial patents filed – November 2007

Technology transferred to NuScale – November 2007

Initial Financing – January 2008

Request to initiate pre-application discussions sent to NRC – January 2008

Signed MOU with Peter Kiewit Company – April 2008

7

Strategic Partner - Kiewit Construction: NuScale / Kiewit MOU signed April 2008

Employee-owned company; $6 billion annual revenue with 120 year history and 16,600 Employees

FORTUNE’s most admired company in the engineering and construction industry in 2007

Major power plant constructor

Major commitment to new nuclear projects based on past nuclear construction experience

Full “one-stop shop” capability

250-acre manufacturing facility in Corpus Christi, Texas

Kiewit Corporate HeadquartersOmaha, NE

8

Nuclear QualityNuclear Quality Activities

10CFR50 Appendix B Program in place

ASME CERTIFICATES – status

• MODULE CONSTRUCTION

Kiewit will be receiving Certificates of Authorization for Division 1 and Division 2 (metallic parts only) for NA, NPT, and NS (July 2008)

• SITE CONSTRUCTION

Nuclear Certificates of Accreditation in place (NA, NPT, NS, CC)

9

Industry Partners

10

Economic Advantages:NuScale controls financial risks and avoids “pinch points”

Controls financial risksNew Capacity matches demand growth

Off-site pre-fabrication and simplicity limit construction risks

Avoids “Pinch Points”The NSSS is prefabricated off-site by domestic manufacturers

Onsite construction force requirements reduced

Many suppliers (pumps, pipes, etc) eliminated by plant simplicity

Forgings for conventional nuclear plants done by Japan Steel Works.

11

Pre-Application Review Objectives

Provide for timely identification of regulatory requirements

Provide for timely, independent assessment of safety and security characteristics

Familiarize NRC staff with the design’s safety case, security approach, and technology base

Inform the development of NRC technical infrastructure: knowledge, analysis tools, data

Provide timely feedback to pre-applicant on:Key safety, design, and licensing issues; and

Technology development program plan

12

Outline for DiscussionDesign Familiarization

Design Safety and Security Approach

Overall Approach to LicensingExisting LWR Licensing Framework

Codes and Methods

Integral Test Facility and Proposed Test Program

Proposed topics and schedule for future pre-application meetings

13

Proposed Pre-Application Schedule

FY2008 FY2009

4Q 1Q 2Q 3Q

1st Meeting● NuScale and Design Introduction ▼

Submit Design DescriptionReport ▼

2nd Meeting● Codes and Methods Topical Report ▼

3rd Meeting● Online Refueling Topical Report● Multi-Module I&C and Operations

Staffing Topical Report

▼

4th Meeting● Multi-Module PRA Topical Report● Severe Accidents Topical Report● Dose Calculations and Emergency

Planning Topical Report

▼

14

Introduction to NuScale Design

Dr. José N. Reyes, Jr.Chief Technical Officer

NuScale Power Inc.

July 24, 2008


Rockville, MD

OutlineSingle Power Module

Multi-Module Plant

Refueling Process

I&C Approach

Multi-Module Control Room

Engineered Safety Features

Expert Panel Review

Severe Accident Mitigation and Prevention

Security and Safeguards Advantages

Conclusions2

Power Module

3

Power Module45 MWe

Simple and Robust DesignIntegrated Reactor Vessel enclosed in an air evacuated Containment Vessel

Immersed in a large pool of water

Located below grade

Utilizes off-the-shelf turbine-generator set

Negatively buoyant (slightly) module with seismic supports on the side (not shown) Multiple fission product barriers

4

Plant CharacteristicsPower Generation Module

• Reactor Type PWR

• Electrical Output 45 MWe

• Steam Generator Number Two independent tube bundles

• Steam Generator Type Vertical, once-through, helical tubes

• Average Steam Generator Tube Length 22.3 m (73.2 ft)

• Steam Generator Tube Number ~1000

• Steam Cycle Superheated

• Turbine Type 3600 rpm, single pressure

• Steam Flow 56.1 kg/s (445,000 lb/hr)

Reactor• Thermal Power 150 MWt

• Reactor Pressure and Core Exit Temperature

P < 10.4 MPa (1500 psig), 575 K (575 F)

• Primary Coolant Mass Flow Rate ~600 kg/s (4.76E6 lb/hr)

• Refueling Intervals 30 months, UO2, 4.95% enriched

5

Standard Materials

Carbon steel containment

Carbon steel vessel, stainless steel lined

Standard PWR-type fuel (half-height) Stainless steel/Inconel steam generators

Stress corrosion cracking issues reduced due to lower operating pressures and temperatures

6

PWR Core Analysis CodesCASMO-4

SIMULATE-3Core Features

24 - 17x17 standard fuel assemblies1.82 m (6 ft) lengthFour control rod drivesFour clusters/driveSixteen assemblies with control rod clusters

Additional Core Design GoalsRapid load following

Core DesignCore Neutronics Optimization

Ø 2.342Ø 2.742Ø 1.50

Ø 1.70

Cross Section B

a a

a a

b b

b bc c d d

c c d d

Ø .811

Ø .203

7

Integrated Reactor VesselImproved Reliability

2.7 m (9 ft) OD , 13.7 m (45 ft) long7.6 cm (3.0 in) carbon steel vessel with internal stainless steel linerReduced operating pressure: 10.4 MPa (< 1500 psig) Natural circulation flow (no pumps) 150 MWth UO2 LWR corePressurizer heaters (not shown) Steam Generator

Two Independent helical coil tube bundles

Two feedwater inlets

Two main steam outletsCore shroud and riserFour CRDMs/16 control rod clustersTwo reactor vent valvesTwo sump valvesOne flange location 8

High Pressure ContainmentEnhanced Safety

Capable of 3.1 MPa (450 psia) Equilibrium pressure between reactor and containment following any LOCA is always below containment design pressure.

Insulating VacuumSignificantly reduces convection heat transfer during normal operation.

No insulation on reactor vessel. ELIMINATES SUMP SCREEN BLOCKAGE ISSUE (GSI-191).

Improves steam condensation rates during a LOCA by eliminating air.

Prevents combustible hydrogen mixture in the unlikely event of a severe accident (i.e., no oxygen).

Eliminates corrosion and humidity problems inside containment.

9

Multiple-Module Complex – Flexible Capacity (12 modules – 540 MWe)

10

Reactor Building/Turbine Generator Buildings Transversal Elevation View

11

Refueling ProcessReactor shutdown using normal feedwater

Decay heat removal transferred to passive DHRS

Containment partially flooded

Module disconnected from all piping and instrumentation

Module is connected to crane and transferred to refueling pool

Lower containment is removed

Lower reactor head is removed

Core is shuffled/reloaded

Replace lower reactor head and containment

Module is transferred back to reactor bay

Module is reconnected

Containment is drained and module is restarted

Refueling Animation

13

I&C ApproachMulti-Module Operation

State-of-the-Art Digital System DesignComplete separation between safety and non-safety systems

Redundant safety actuation systems

Upfront PC-based testing to inform control room design

Incorporation of lessons learned/data from related industries (e.g., air traffic controllers)

Human-Machine Interface Verification and ValidationSingle module testing

Full-scale integrated simulator testing

14

Multi-Module Control Room

15

Engineered Safety FeaturesHigh Pressure Containment Vessel

Shutdown Accumulator System (SAS) Passive Safety Systems

Decay Heat Removal System (DHRS) Containment Heat Removal System (CHRS)

Severe Accident Mitigation and Prevention Design Features

16

Decay Heat Removal System (DHRS)

Two independent trains of emergency feedwater to the steam generator tube bundles.

Water is drawn from the containment cooling pool through a sump screen.

Steam is vented through spargers and condensed in the pool.

Feedwater Accumulators provide initial feed flow while DHRS transitions to natural circulation flow.

Pool provides a 3 day cooling supply for decay heat removal.

17

Containment Heat Removal System (CHRS)

Provides a means of removing core decay heat and limits containment pressure by:

Steam Condensation

Convective Heat Transfer

Heat Conduction

Sump Recirculation

Reactor Vessel steam is vented through the reactor vent valves (flow limiter).Steam condenses on containment.Condensate collects in lower containment region (sump).Sump valves open to provide recirculation path through the core. 18

Event Response Logic

XSGTR – 2 Tube bundle

Unaffected Train

SGTR – 1 Tube bundle

X

X

X

X

DHRS2

1. Initiating Events cause Reactor Scram

2. Two independent trains each capable of 7% decay heat removal.

XLOCA w/o SGTR

MSLB

Station Blackout

Loss of Feedwater

CHRS2INITIATING EVENT1

19

Expert Panel ReviewJune 2-3, 2008, a panel of experts convened to develop a Thermal-Hydraulics/Neutronics Phenomena Identification and Ranking Table (PIRT) for the NuScale module:

Graham Wallis, Creare (Panel Chairman) Mujid Kazimi, MIT

Larry Hochreiter, Penn State

Kord Smith, Studsvik Scanpower

Brent Boyack, LANL retired

Jose Reyes, NuScale Power, OSU20

Preliminary Panel ResultsLarge-break LOCA eliminated by design

In other words, no large-break LOCA phenomena

Since all water “lost” out of the primary system can be recovered by opening the sump recirculation valves, it is impossible to uncover the core during design bases LOCAs

Therefore even a small-break LOCA does not challenge the safety of the reactor

21

Severe Accident Mitigation and Prevention

Reduced source term due to modularization and additional fission product barriers

No need for combustible gas control in containment (containment inerted) No molten concrete coolant interactions

Reliable and redundant reactor depressurization system (no high-pressure melt ejection)

22

Additional Fission Product Barriers

NOT TO SCALE

Fuel Pellet and Cladding

Reactor Vessel

Containment

Containment Cooling Pool Water

Containment Pool Structure

Biological Shield

Reactor Building

23

Security and Safeguards Advantages

Safety maintained without external power

Below-grade

Power Module (NSSS and Containment)

Control Room

Spent Fuel Pool

Low profile building

Containment pool Impact Shield for aircraft

24

ConclusionsSimple and Robust Design

Maximizes safety and security through use of passive systems, modularity, and multiple fission product barriers

Natural circulation eliminates failure modes and need for pumps

Integrated power module eliminates unnecessary piping and improves reliability

Large-break LOCAs eliminated by design and small-break LOCAs do not challenge the safety of the plant

Probability of post-DCD design revisions are significantly reduced due to simplicity of the design

The NuScale design is based on decades of LWR experience and incorporates numerous innovative safety and security enhancements 25

Approach to Licensing


NuScale Power Inc.

July 24, 2008


Rockville, MD

Sufficiency of Existing NRC Licensing Framework

Codes and Methods

Testing Program

Outline

2

Approximately 95% of the regulatory basis for NRC design review of a multi-module NuScale plant currently exists. Of the 255 sections in the SRP:

217 are directly applicable without modification25 do not apply because

They relate to BWR designsThey apply to components that have been eliminated in the NuScale design

13 pose issues that require additional discussion All relate to multi-module operation

Applicability of NRC Standard Review Plan (NUREG-0800)

3

Relevant SRP Multi-Module Sections(Chapters 7, 13, 14, 18, and 19)

TitleSection

Probabilistic Risk Assessment19

Human Factors Engineering18

Emergency Planning: Inspections, Tests, Analyses, and Acceptance Criteria14.3.10

Human Factors Engineering: Inspections, Tests, Analyses, and Acceptance Criteria14.3.9

Instrumentation and Controls: Inspections, Tests, Analyses, and Acceptance Criteria14.3.5

Operating and Emergency Operating Procedures13.5.2.1

Operational Programs13.4

Emergency Planning13.3

Non-Licensed Plant Staff Training13.2.2

Reactor Operator Requalification; Reactor Operator Training13.2.1

Data Communication Systems7.9

Diverse Instrumentation and Control Systems7.8

Controls Systems7.7

4

Topical Areas to Address Multi-Module Issues

Multi-Module I&C and Operator Staffing

Multi-Module PRA

Dose Calculations and Emergency Planning

Online Refueling

5

Sufficiency of Existing Licensing Framework

Technical basis for licensing a single NuScale module is well established and current NRC regulations are sufficient

NuScale plant technology is based on decades of light-water reactor experience.

NuScale will rely on NRC-sponsored safety analysis codes assessed against design-specific, large-scale integral and separate effects data for transient and accident analysis.

Sections of the SRP which do not address multi-module operations will be addressed through Topical Reports with exemption requests, as necessary.

6

NuScale design is also responsive to the NRC's updated Draft Advanced Reactor Policy Statement...

Highly reliable and less complex shutdown and decay heat removal systemsLonger time constants and better instrumentation before challenging safety systemsSafety systems that reduce operator actions and equipment subjected to severe environment conditionsMinimize potential for severe accidents and their consequencesImproved reliability in balance of plantEasily maintainable equipment and components

7

...and meets all policy statement goals.

Reduce radiation exposure to plant personnelIncorporate defense-in-depth philosophy by maintaining multiple barriersRely on proven technologyInclude considerations for safety and security requirements during design phasePrevent simultaneous loss of containment integrity and ability to maintain core cooling (e.g., aircraft impact) Prevent loss of spent fuel pool integrity as a result of aircraft impact

8

Codes and MethodsDemonstrate adequacy of NuScale codes and methods for analyzing transients and accidents

Use of well-established computer codes and methods based on decades of LWR experience

Conduct design-specific integral and separate effects tests to validate safety system performance and evaluation models

Address uncertainties in a rigorous and integrated manner

9

Safety Analysis Code EvaluationDESIGN BASES ACCIDENT ANALYSIS

● Fuels Analysis FRAPCON-3, FRAPTRAN, COBRA-TF

● Core Analysis CASMO-5, SIMULATE-3

● ECCS Performance RELAP53.3, RELAP5-3D, TRACE5.0, Star-CD

● Containment Performance RELAP53.3, RELAP5-3D, TRACE5.0, Star-CD

● Control Room Habitability RADTRAD3.03

● Offsite Dose RADTRAD3.03

SEVERE ACCIDENT ANALYSIS

● Fuel Integrity FRAPCON-3, FRAPTRAN, SCDAP/RELAP3.3, MELCOR1.86

● Reactor Vessel Integrity SCDAP/RELAP3.3, MELCOR1.86, ANSYS

● Containment Integrity MELCOR1.86, ANSYS

● Control Room Habitability RADTRAD3.03

● Offsite Dose RADTRAD3.03

RADIATION PROTECTION MCNP5.1

CRITICALITY SAFETY MCNP5.1

PRA SAPHIRE7.2710

Proposed Certification TestingIntegral System Test Categories

SBLOCA (inadvertent opening of reactor vent valves and sump valves, CVCS line break, etc.) Secondary side ruptures (main steam/feed line breaks) Station blackout

Full Scale Helical Coil Steam Generator Tests (reduced tubes)

Confirmation of heat transfer and stability performance

Full Scale Valve Tests

11

NuScale Integral Test FacilityStainless steel integral system test facility operating at full system pressure and temperature

Reactor Vessel

Electrically heated rod bundle

Core Shroud with Riser

Pressurizer

Sump Recirculation Valves

Helical Coil Steam Generator

Variable Speed Feedwater Pump

Containment Vessel

Containment Cooling Pool

Instrumentation 12

Integral System Test FacilityA scaling analysis was used to guide the design, construction and operation of a 1/3-scale integral system test facility for the original MASLWR designNuScale has exclusive use of the test facilityNuScale is modifying the facility to incorporate design improvementsFacility can be used to:

Evaluate design improvementsConduct integral system tests for NRC certification

OSU has significant testing capabilityPerformed DOE and NRC certification tests for the AP600 and AP1000 designs.10 CFR 50 Appendix B, NQA-1, 10 CFR 21

13

Scaling Ratios

1:1Temperature Ratio

1:1Pressure Ratio

1:3.1Velocity Ratio

1:1Time Ratio

1:1Active Heat Transfer Wall Thickness Ratio

1:254.7Active Heat Transfer Area Ratio

1:254.7Power Ratio

1:254.7Volume Ratio

1:82Cross-sectional Area Ratio

1:3.1Length Ratio

14

Integrated Reactor Test Vessel

Pressurizer

PZR Steam Drum

SG Helical Coils

Core Shroud

Riser

Flange

Pressure Vessel

Core Heaters

15

Containment and Cooling Pool

Trace Heated High Pressure Containment

Containment Cooling Pool

Containment Heat Transfer Plate

16

SBLOCA Transient Phases

Phase 1: Blowdown PhaseBegins with the opening of the break and ends with the reactor vent valve (RVV) initiation

Phase 2: RVV OperationBegins with the opening of the reactor vent valve and ends when the containment and reactor system pressures are equalized

Phase 3 - Long Term CoolingBegins with the equalization of the containment and reactor system pressures and ends when stable cooling is established via opening of the sump recirculation valves

17

Integral Testing Schedule (Oregon State University)

Facility refurbishment and QA program setup – August 2008

Characterization and shakedown testing – Sept. 2008Includes integrated steam generator performance testing

Single blowdown test – Nov. 2008

Additional facility modifications – Feb. 2009

Phase I Integral Testing (~5 tests) – July 2009

Phase II Integral Testing (~10 tests) – Dec. 2009

20

ConclusionsTechnical basis for licensing a single NuScale module is well established and current NRC regulations are sufficient.

Sections of the SRP which do not address multi-module operations will be addressed through Topical Reports with exemption requests, as necessary.

Adequacy of NuScale codes and methods for analyzing transients and accidents will be demonstrated.

Integral System Testing

Full height, reduced tube, helical coil steam generator tests

Full scale valve tests

21

Identification of ProposedPre-Application Topical Areas


NuScale Power Inc.

July 24, 2008


Rockville, MD

OutlineOnline RefuelingMulti-Module Digital I&C and Operations StaffingMulti-Module PRASevere AccidentsDose Calculations and Emergency PlanningCodes and Methods

2

Online RefuelingUnique NuScale Design Features

Individual modules can be taken offline and refueled while other modules continue to operateContainment and reactor are moved to refueling bay

Key Elements to Address in Topical ReportRefueling process and its applicability to existing regulationsDecay heat removal system performanceStaffing and training

3

Multi-Module Digital I&C and Operations Staffing

Unique NuScale Design FeaturesOne operator for multiple units

Early notification of approach to safety setpoints

Automatic safety system actuation

Key Elements to Address in Topical ReportDescription of technical bases for 50.54(m) exemption request (NUREG-1791) Integration of human performance data

Simulator design and testing program

4

Multi-Module PRAUnique NuScale Design Features

Multiple, physically separate 150 MWt power modules

Passive safety systems

Additional fission product barriers

Common cause failures leading to core damage reduced by physical separation and no sharing of safety systems

Key Elements to Address in Topical ReportMulti-module common mode failures (e.g., loss-of-offsite power, fires, floods, etc.) Dominant accident sequences (Level I and II PRA results) for single and multiple units

Passive safety system performance/reliability

Data quality and applicability 5

Severe AccidentsUnique NuScale Design Features

Low power core

No reactor vessel insulation

Evacuated steel containment


Key Elements to Address in Topical ReportCompliance to existing NRC regulations and Commission policy statements

In-vessel retention/ex-vessel cooling strategy

No need for combustible gas control in containment

6

Dose Calculations and Emergency Planning

Unique NuScale Design FeaturesSmall source term, ~5% of a 3000 MWt PWR

Additional fission product barriers (e.g., Containment Cooling Pool water, impact/biological shield, reactor building) Severe accidents leading to a larger source term would require multiple simultaneous failures in more than one module

Key Elements to Address in Topical ReportApplicability of existing source term guidance

Determination of offsite doses under severe conditions to be used for emergency planning

Level II Multi-Module PRA results 7

Codes and MethodsUnique NuScale Design Features

Small, half-height core

Natural circulation under normal operation


Key Elements to Address in Topical ReportSafety analysis calculational framework

Experimental programs and applicability of existing LWR benchmark data

Selection of computer codes

Verification and validation plans

Phenomenon Identification and Ranking Table Results8

Proposed Pre-Application Schedule

FY2008 FY2009

4Q 1Q 2Q 3Q

1st Meeting● NuScale and Design Introduction ▼

Submit Design DescriptionReport ▼

2nd Meeting● Codes and Methods Topical Report ▼

3rd Meeting● Online Refueling Topical Report● Multi-Module I&C and Operations

Staffing Topical Report

▼4th Meeting

● Multi-Module PRA Topical Report● Severe Accidents Topical Report● Dose Calculations and Emergency

Planning Topical Report

▼

9

Legal Issues

James Curtiss

July 24, 2008


Rockville, MD

Certification of Multi-Modules

Assessment of Part 171 Fees

Outline

2

DC application will seek certification for a multi-module plant

Issues associated with the subsequent addition of modules and multi-module operation will be resolved in design certification (e.g., shared/ common systems, construction during operation)

COL applicant will reference the multi-module design certification

Site-specific issues will be focus of COL review

Single COL would permit the addition of modules and operation of the multi-module plant referenced in design certification

COL holder would build/operate modules as demand necessitates

40-year period for individual modules would begin upon finding by Commission that acceptance criteria are met under 52.103(g) for individual modules

Certification of Multi-Modules

3

As a single COL would be issued referencing the multi-module design certification, a single annual fee would be assessed under Part 171 for a COL review.

Assessment of Part 171 Fees

4

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