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Japanese Codes and Standards

to Support Long Term Operation

of Nuclear Power Plant

23-27 October, PLiM 2017

Masahito Mochizuki

Osaka UniversityDivision of Materials and Manufacturing Science,

Graduate School of Engineering

IAEA-CN-246-038

Contents

1. Outline of Japanese Codes and Standards to support

Long Term Operation of Nuclear Power Plant

2. Main Codes and Standards related to Plant Life

2-1. Codes and Standards for Design/Construction

2-2. Codes and Standards for Operating & Maintenance

2-3. Codes and Standards for Decommissioning

3. Current Activities and Future Programme

1

� Establish voluntary codes/standards as the detailed specifications to assure safety of

NPP facilities under the principle of “fairness”, “equitableness” and “openness” through

a process with assured transparency, including invitation of public comments and then

go through the endorsement by NRA (Nuclear Regulation Authority) so that they can be

applied to actual plants.

� Incorporate the latest technologies and knowledge in a prompt manner and formulate

standardized codes and standards to facilitate the application to actual plants.

� Promote licensee’s voluntary efforts to enhance safety.

Importance of voluntary codes/standards

Japanese Voluntary Codes/Standards

related to Nuclear Power Plant

The Japanese academic societies and associations* have established a number of voluntary codes and standards, which are intended to support the nuclear power plant throughout its plant life, and those codes and standards have been applied to actual plants.

�Atomic Energy Society of Japan,

�Japanese Society of Mechanical Engineers

� Japan Electric Association, etc.

*Major Academic Societies and Associates

Development of Codes/Standards to Support a NPP throughout Life

Safety management/quality assurance activities

Maintenance management activities

Safety (system) design/evaluation

Design considering structural strength of materials (including seismic design)

Operation management, fuel management, radiation management, etc.

Periodic safety review

Plant life

management

Planning,

implementation and

management of

decommissioning

Design/construction Operation and maintenance Decommissioning

Laws & ordinances, regulations, NRA Guidelines, etc.

Consensus Codes and Standards

Establishment/revision

2

Application

AESJ JSME JEA Other academic societies

Safety Management System Code (Quality Assurance Code) （JEAC4111）

Maintenance Management Code （JEAC4209）

*Revision of related codes/standards is being studied to cope with the new inspection system

covering the entire scope of utilities activities.

Decommissioning

Plan/Implementation

Individual

equipment

（Example）

Reactor Vessel：

Core Internal：

Pipe：

Overview of Codes/Standards Related to Nuclear Power Plants

throughout Life ［Design～Long-term Operation of LWR］

Design/Construction In-service period Decommissioning

Management

system

Human factor

Guidelines/rules related to safety design (guidelines on protective measures against natural disaster, flooding, fire, etc.)

Rules on Materials for Nuclear Facilities （JSME S NJ-1）

Rules on Design and Construction for Nuclear Power Plants （JSME S NC-1）

・Separate guidelines for individual items, including high cycle fatigue and fluid force vibration

Technical Code on Aseismic Design （JEAC4601）, Technical Guideline on Tsunami Resistance Design （JEAG4629）

Rules on Welding for Nuclear Power Plants （JSME S NB-1）

Rules on Fitness-for-Service for NPPs （JSME S NA-1）

・Guidelines on NDT (UT, ET, etc.), CV-LRT, etc.

・Environmental Fatigue Evaluation Methods

Technical Guidelines on RV Irradiation Embrittlement Management and

Evaluation （JEAC4201, JEAC4206）

［JANSI］ Core Internals Inspection/Evaluation Guidelines

・Rules on Pipe Wall Thinning Management for PWR(BWR) Power Plants , etc.

PSR＋ Guideline

Code on Ageing Management

Syste

m/s

tru

ctu

re

/ co

mp

on

en

t

De

sig

n/

ma

nu

factu

ring

・Inspection

・evaluation

・repair

・Ageing

・monitoring/

・assessment

Plant life management


Development of

guidelines on

safety evaluation

is being studied

based on the

guidelines on

planning of

decommissioning.

* Discussion over the revision of related codes/standards are being pursued on a continuous basis considering lessons learned from the Fukushima

Daiichi accident and severe accident management measures, utilizing risk information, and incorporating, operating experience and updated

test/research results.

3

○ Japan Society of Mechanical Engineers (Committee on Code for Nuclear

Power Generation Facilities) ⇒ JSME CodeJSME Codes mainly cover the areas related to the assurance of integrity of structures and functions of mechanical

facilities, including material, design, fabrication, test/inspection, maintenance management and dismantling.

【Examples of JSME Codes related to LTO】

“Rules on Materials for Nuclear Facilities”, “Rules on Design and Construction for Nuclear Power Plants”,

“Rules on Welding for Nuclear Power Plants” , “Rules on Fitness-for-Service for Nuclear Power Plants”, etc.

○ Japan Electric Association (Nuclear Standards Committee) ⇒ JEAC/JEAGJEAC/JEAG provide for the rules on maintenance management and safety management of electric works for nuclear

power generation, covering a wide technical areas, including operation and seismic design of nuclear power plants.

【Examples of JEAC/JEAG related to LTO】

“Technical Guideline for Seismic Design of Nuclear Power Plants”, Code on Management System for Nuclear Safety”,

“Code for Maintenance Management of Nuclear Power Plants”,

“Methods for surveillance tests of structural materials of nuclear reactors”, etc.

◎ Association for Academic Societies’ Codes and Standards for Nuclear Power Generation

A forum for information exchange and discussion about role sharing between the above three academic societies

regarding the establishment of codes and standards related to nuclear power plants and other nuclear facilities.

Major Academic Societies and Their Codes/Standards 4

○ Atomic Energy Society of Japan (Standard Subcommittee) ⇒ AESJ CodeAESJ Codes cover a wide range of technologies, including site selection of a nuclear facility, fundamental design, system

design, maintenance, decommissioning, reprocessing, treatment and disposal of radioactive waste, use of radiation.

【Examples of AESJ Codes related to LTO】

“Code on Implementation and Review of Nuclear Power Plant Ageing Management Programs”,

“Proactive Safety Review (PSR+) Guideline for Continuous Improvement of Nuclear Power Plants”

“Planning and Implementation of Decommissioning of Nuclear Facilities”, etc.

Establishment and Revision of Consensus Codes

and Standards to Cope with Recent Situations

�To take lessons learned from the Fukushima Daiichi accident

Establishment and revision of codes and standards to meet

the new NRA requirements

・Enhances the design basis for natural phenomena

(earthquake, tsunami, tornado, etc.) ,strengthens fire

protection measures, etc.

・Considers severe accident management measures

�To incorporate the latest technical knowledge in Japan and

abroad as well as IAEA guides

・Incorporates experience with maintenance management at

Japan’s NPPs and information from IAEA I-GALL into the

Code on Plant Life Management

・Introduces the PSR+ Standard

5

Conventional Regulatory Requirements

New Regulatory Requirements

(since Jul. 2013)

Severe accident measures(Licensees‘ self-imposed

safety measures)

New

Establishment and Revision of Related Codes/Standards

6

Natural phenomena

Reliability of power supply

Function of other SCCs

Ultimate heat sink

Fire

Seismic/Tsunami resistance

Prevention of core damage

Seismic/Tsunami

resistance

Ultimate heat sink

Fire

Function of other SCCs

Internal flooding

Natural phenomena

Prevention of CV failure

Suppression of radioactive materials dispersal

Specialized Safety Facility

Reliability of power supplyReinforced

or

New

Reinforced

Takahama NPP

Reinforcement of Safety Measures

Recent Major Developments in Setting and Revising Codes/Standards (1/2)

○ Atomic Energy Society of Japan

� Revision of Implementation Standard Concerning Severe Accident

Management （revised in 2016, under further discussion）

� Establishment of Implementation Standard Concerning Risk–Informed

Decision Making (RIDM) (under discussion）

� AESJ has been continuously working on the utilization of risk information

in Proactive Safety Review (PSR+)and severe accident management.

� Revision of the Code on Implementation and Review of Nuclear Power

Plant Ageing Management Programs (subject to continuous update)

○ Japan Society of Mechanical Engineers

� Establishment of the Guideline for Evaluation of Impact Loads by

Tornado Missiles on Structural Integrity （under discussion）

� Revision of Rules on Design & Construction, Materials, Welding, and

Fitness-for-Service for NPPs (subject to continuous update)

7

8

○ Japan Electric Association

� Establishment of codes/guidelines regarding natural phenomena etc.

- JEAC4601-2015 “Technical Code for Seismic Design of NPPs/JEAG4601-2015”

- JEAC4629-2014 “Technical Code for Tsunami Resistance Design of NPPs”

- JEAG4630-2016 “Technical Guideline for Flood Prevention Facilities ”

- JEAG4625-2015 “Technical Guideline for Evaluation of Volcanic Hazards”

（under further discussion）

� Establishment of codes/guidelines regarding design of Severe Accident Instrumentation,

emergency power supplies, environmental qualification, digital safety protection system,

verification of environmental qualification of electric equipment and instrumentation ,

human interfaces, etc. （under further discussion）

� Revision of codes/guidelines regarding maintenance

- JEAC4209-2016 “Code for Maintenance Management of NPPs/JEAG4210-2016”

- JEAC4207-2016 “Code for Ultrasonic Testing during In-Service Inspection of

LWR Plant Components”

� Establishment of

- JEAC4111-2013 “Code on Management System for Nuclear Safety”

Recent Major Developments in Setting and Revising Codes/Standards (2/2)





Individual

equipment

（Example）

Reactor Vessel：

Core Internal：

Pipe：




Management

system

Human factor














PSR＋ Guideline


Syste

m/s

tru

ctu

re

/ co

mp

on

en

t

De

sig

n/

ma

nu

factu

ring

・Inspection

・evaluation

・repair

・Ageing

・monitoring/

・assessment



Development of

guidelines on

safety evaluation

is being studied

based on the

guidelines on

planning of

decommissioning.




9

Decommissioning

Plan/Implementation

JEAG 4625 “Technical Guideline for Evaluation of Volcanic Hazards”

In Design/ Construction stage

JSME S NJ1 “Rules on Materials for Nuclear Facilities”

JSME S NC1 “Rules on Design and Construction for Nuclear Power Plants”

JSME S 012 Guidelines on the evaluation of fluid force vibration for cylindrical structure inside pipe

JSME S 017 Guidelines on the evaluation of high cycle thermal fatigue

JSME S NB1 “Rules on Welding for Nuclear Power Plants ”provide for the requirements for the welding process according to the class of components as well as

the qualification of welders and welding techniques

JSME S NE1 “Rules on Design of Concrete Containment Vessels”

JEAC 4601 “Technical Code for Seismic Design of Nuclear Power Plants”

JEAC 4111 “Code on Management System for Nuclear Safety”

JEAC 4629 “Technical Code for Tsunami Resistance Design of Nuclear Power Plants”

JSME

Code

JEAC/

JAEG

� Codes/Standards related to material, structural strength design, earthquake resistance design, welding, etc.

10

11

12

Discussion over evaluation of safety of NPP facilities against tsunami was started in FY2012, and, as a result, JEAC 4629 “Technical Code for Tsunami Resistance Design of NPPs”, which includes basic requirements and methods of tsunami resistance design, was published in September 2014.In future, it is planned to develop a draft revision by adding detailed specifications about tsunami resistance design through studies on the design considering the impact of wave power and collision of drifting debris.

Outline of JEAC4629-2014

“Technical Code for Tsunami Resistance Design of NPPs”

Contents of JEAC4629

Chap. 1 Definition of basic principles (basic design principle, importance

classification, policy on protection, assumed events, etc.)

Chap. 2 Procedure of designing tsunami resistance

Chap. 3 Impacts of tsunami

Defines the methods for evaluating the impacts, which are used

as the design input (Analysis of the impacts of tsunami run-ups,

wave power, collision of drifting debris, etc.)

Chap. 4 Tsunami resistance design of tsunami protection facilities and

flood prevention equipment

Chap. 5 Tsunami resistance design of components/electric equipment

(Performance targets of individual facilities, external forces to be

applied, loads to be considered, etc.)

Chap. 6 Tsunami resistance design against tsunami-induced events

Chap 7. Evaluation of tsunami resistance performance of reactor facilities

Chap. 8 Detection of tsunamis and operation management

Reference material: examples of damages caused by tsunamisTakahama NPP

11

Outline of JEAG4625-2015“Technical Guideline for Evaluation of Volcanic Hazards” (1/2)� 2009: The guideline for evaluation of volcanic hazards to be considered in site selection process

was established. : sets the concept for evaluation of target volcanoes and the criteria for setting

the extent of volcanic effects

� Revision following the Fukushima Daiichi accident

2014: Defined the guideline for evaluation of volcanic hazards on mechanical and electric

equipment to be considered in the detailed design stage

※Sorting out the modes of impact by ash fall considering structural characteristics of facilities

� 2015: Added items to be considered for the facilities to deal with specific severe accidents

� Future plan ⇒ discussion over “uncertainties specific to volcanic phenomena”

Contents of the guideline

Chap. 1 Basic principles

Chap. 2 Investigation and evaluation of

volcanoes and volcanic phenomena

Chap. 3 Evaluation of volcanic effects on

mechanical and electric equipment

Flow of evaluationStep 1: Select volcanoes subject to the evaluation by

referring to previous literature↓

Step 2: Select volcanoes which may erupt during servicebased on the literature survey, topographic survey, geological survey, etc.

↓Step 3: Evaluate potential effects of volcanic phenomena

on a NPP・Volcanic ash (including pumice)・Volcanic gas

↓Step 4: Evaluate the feasibility to take corresponding

action in the detailed design and operation stages

Fume

Volcanic gas

Volcanic mudflow

Debris avalanche

Pyroclastic flow/surge

Bomb

Ash fall

Lava flow

12

Basic principles to be considered in the design and operation of plant facilities

① A plant should be transferred from a hot shutdown state to a cold shutdown state,

and the cold shutdown state should be maintained after the plant is shut down safely.

② Cooling functions of the spent fuel storage pool should be maintained.

・Examples of components , including outdoor tanks, seawater pumps, heat exchangers,

building ventilation systems, for which design considerations are taken

・Examples of specific design features for facilities to deal with severe accidents and other disasters

【Examples of facilities that may be subject to the impact of ash falls, etc. (ABWR)】

Screen

Condensate water

storage tank

(deposition)

Transformer

(insulation

degradation)

Reactor building

(deposition)

Main vent stack

(clogging)

Turbine building

(deposition)Filter

(clogging)

Component cooling

seawater pump

(intrusion)

Light oil tank

(deposition)Turbine

To

ma

in sta

ck

Generator

OG system

Condenser

Intake

air

Exhaust

airReactor

CV

SFP

Emergency diesel generator

(intrusion)

13Outline of JEAG4625-2015“Technical Guideline for Evaluation of Volcanic Hazards” (2/2)





Individual

equipment

（Example）

Reactor Vessel：

Core Internal：

Pipe：




Management

system

Human factor














PSR＋ Guideline


Syste

m/s

tru

ctu

re

/ co

mp

on

en

t

De

sig

n/

ma

nu

factu

ring

・Inspection

・evaluation

・repair

・Ageing

・monitoring/

・assessment



Development of

guidelines on

safety evaluation

is being studied

based on the

guidelines on

planning of

decommissioning.




14

Decommissioning

Plan/Implementation

JEAC 4629 “Technical Code for Tsunami Resistance Design of NPPs”

JSME S NJ1 “Rules on Materials for Nuclear Facilities”

JSME S NC1 “Rules on Design and Construction for Nuclear Power Plants”

JSME S NB1 “Rules on Welding”

JEAC 4601 “Technical Code for Seismic Design of Nuclear Power Plants”

JEAC 4111 “Code on Management System for Nuclear Safety”provides for the requirements for quality assurance, including the establishment of

quality management activities.

JSME

Code

JEAC/

JEAG

15

� Codes/Standards related to O&Mm activities

JSME S NA1 “Rules on Fitness-for-Service for Nuclear Power Plants”provide for the detailed standards for in-service inspection of components, flaw evaluation

and repair techniques applicable to components.

JEAC 4209 “Code for Maintenance at Nuclear Power Plants”The Japanese utilities establish their maintenance programs for NPPs based on this

code and perform maintenance management activities according to the programs.

� Codes/Standards related to in-service inspection, evaluation, repair, etc.

During Operation & Maintenance stages ( 1/2 )

17

Guidelines on NDT (UT, ET, etc.), CV-LRT, etc.

JEAC/

JEAG

� Codes/Standards related to ageing management evaluation and monitoring

Technical Guidelines on Irradiation Embrittlement of Reactor Vessel

JEAC4201 :the method of surveillance tests for structural material shock

JEAC4206 :the method of integrity assessment associated with fracture toughness,

including pressurized thermal

JSME S CA1/NH1/NG1 Rules on Pipe Wall Thinning Management

Japan Nuclear

Safety Institute

JSME S NF1 Environmental Fatigue Evaluation Methods

“Inspection and Evaluation Guideline for Core Internals”.

16During Operation & Maintenance stages ( 2/2 )

JEAC/

JEAG

JSME

Code

18

20

� Codes/Standards related to Periodic Safety Review,

PLM(Ageing Management Technical Evaluation), etc.

AESJ

Code

AESJ “Proactive Safety Review (PSR+) Guideline for Continuous

Improvement on Nuclear Power Plants”

AESJ “Code on Implementation and Review of Nuclear Power Plant

Ageing Management Program”

21

22

19

It specifies the requirements for the utility to establish the Management System for Nuclear Safety.

It also incorporates requirements of leadership to enhance nuclear safety and encouragement of

safety culture so that the utility can maintain and enhance Nuclear Safety through PDCA cycle.

Chap. 1 :Purpose, Chap. 2 :Application boundary, Chap 3 :Definition

Chap. 4 :Management System for Nuclear Safety

Chap. 5 :Responsibility of Management

5.5.3 responsible person of process

encouragement of activities on safety culture

Chap. 6 :Operation Management of Resources

Chap. 7 :Plan and Implementation

Chap. 8 :Evaluation and Improvement

Chap. 9 :Safety Culture and Leadership for Nuclear Safety

9.1 Leadership for nuclear safety

9.2 Continuous improvement of safety culture

9.3 Assessment for safety culture and leadership for nuclear safety

� Clearly defines utility’s efforts to enhance Nuclear Safety

� Incorporates new regulatory requirements (enforced in July 2013)

� Utilizes overseas knowledge : IAEA Code DS456(GSR Part2)

【Main amendments】

JEAC4111-2013

17Outline of JEAC4111-2013 “Code on Management System for Nuclear Safety”

Methods for Evaluation of Neutron Irradiation Embrittlement in RPV

Japan Electric Association Codes (JEAC)

� JEAC4201-2007 (Supplement 2013): Method of Surveillance tests for structural materials of nuclear reactors Specifies the surveillance test methods to monitor changes in the mechanical properties of reactor vessel steel

materials due to neutron irradiation, and the equation to predict a shift in the reference temperature

* The method to predict a shift in reference temperature has incorporated the recent mechanistic understandings on

neutron irradiation embrittlement.

� JEAC4206-2016: Verification method of fracture toughness for in-service reactor pressure vesselSpecifies the method to verify fracture toughness of PRV against non-ductile fracture and ductile fracture, and the

evaluation method of PTS for RWR.

� JAEC4216-2015 :Test method for determination of reference temperature, To, of ferritic steels

Specifies the fracture toughness test method for determining the “To” of the master curve. Miniature compact test

specimens (0.16 TCT) has been allowed in the latest revision.

( ) ( )22

MDSCNDTTTRT ∆+∆=∆

SCT∆ ：Increase in transition temperature due to formation

of solute atom clusters

MDT∆ ：Increase in transition temperature due to damage to matrices

【Prevention of non-

ductile fracture】

【Prevention of ductile

fracture】

Requirements for

temperature/pressure

Evaluation of PTS

Incorporating master curve based

RTNDT(RTTo)

Considering the effects of cladding

welds and WPS and revision of

fracture toughness curvePrediction of upper shelf

energy

18

JSME S NG1 Rules on Pipe Wall Thinning Management

for PWR Power Plants

Rules for PWR plant pipe wall thinning management, that specify management method

on pipe wall thickness measurement, are formulated by JSME.

Rules indicate piping systems to be manage and process of thinning management for

FAC and LDI.

[Management process]

Examination Planning- Identify examination scope

- Planning examination schedule

Examination :Normal / Detail measure

Evaluation : Thinning rate, remaining life

Maintenance

Repair/Replacement

Major Revised contents in 2016Rev.

� Revise examination area (piping systems and parts) from latest knowledge

� Specify second measurement schedule

� Enhance management requirement for piping connecting points

Refer to JSME論文集(B)79-808(2013-12),Hirai/Nakamura/Amano

19

Development of Guideline for Inspection of Core Internals

� The Japan Nuclear Safety Institute has set the guideline describing the methods of inspection and repair

techniques for nuclear industry’s voluntary safety improvement efforts.

� The guideline also considers the effects of ageing degradation on the structural integrity of core internals.

� The results of the evaluation of inspections are incorporated into JSME Rules on Fitness-for-Service for NPPs.

【Guideline for inspections of PWR core internals】 The guideline specifies;

・components/parts subject to inspection

and scope of inspection

・methods, timing of start, frequencies of

inspections

・evaluation of inspection results

20

Locking weld

Baffle former bolt

Former plate

Baffle plate

Core barrel

Components subject to detailed inspection(also to general inspection)[Flow ①]Control rod cluster guide tubesBaffle former boltsRV bottom mounted instrumentation nozzles

Components subject to general inspection(also to general inspection)[Flow ②] [Flow ③]Upper core support plate Thermocouple lead tube Flexture pin Upper core support column Upper core plate guide pin Thermal shieldUpper core plate Fuel assembly guide pin Irradiation testCore barrel Baffle plate specimen guide tubeLower core plate Former plateLower core support column Barrel former boltLower core support plate Radial keyRV positioning pin Crevice insertSpray nozzle BMI guide tubeHold ring Secondary core support columnSupport pin RV head nozzleLevel meter

Figure 1 Flow of selection of components/parts subject to general inspection

Core Internal

Does it have safety

functions?

No

Out of scope

Not applicable

Is there any significant

degradation mode?

No: Flow ②

Yes

Yes[Flow ①]

Detailed inspection

(general inspection, as well)General inspection

Applicable:

Flow ③*Those components that have

undergone preventive maintenance has been performed, are classified

in “no significant degradation”.

Is it critical for continuous

operation and asset

management?

Although periodic safety reviews (PSRs) had been performed in Japan, they did not help prevent

the Fukushima Daiichi NPS accident.

Conventional PSRs, which focused on reviewing operational

safety activities, did not lead to effective countermeasures.

Proactive Safety Review (PSR+) Guidelines 2015were developed by the Atomic Energy Society of Japan with reference to IAEA’s PSR Guidelines

(SSG-25), which allow us to predict the future conditions of the plants and to proactively identify

safety enhancement measures in advance.

Overview of PSR+

Safety factor review Overall evaluation

Review the following 14 factors based on the guidelines:

(1) Plant design, (2) Current status of safety significant SSC, (3) Performance

assurance of equipment, (4) Ageing degradation, (5) Deterministic safety

analysis, (6) Probabilistic risk evaluation, (7) Hazard analysis, (8) Actual

safety records, (9) Utilization of experiences in other plants and research

findings, (10) Organization/management system/safety culture, (11)

Procedures, (12) Human factors, (13) Emergency plans, (14) Environmental

impacts of radioactive materials

Extract adequate and feasible safety enhancement measures

↓Confirm safety of future plant operation

↓Develop action plan for safety enhancement measures

Evaluation of Safety Enhancement and Continuous Improvement Proactive Safety Review (PSR+)

21

- The results of the safety enhancement evaluation must be submitted to the Nuclear Regulation Authority in Japan.

- We plan to effectively use PSR+ in the long-term review included in the safety enhancement evaluation which should be submitted in the future.

27

“Code on Implementation and Review of Nuclear Power Plant Ageing Management Programs” by AESJ

Atomic Energy Society of Japan (AESJ) has established the “Code on Implementation and

Review of Nuclear Power Plant Ageing Management Programs” in 2008 and continuously

updated.

Structure of the Code

○ Main text (Chapters 1~9)○ Attachments

A： Summary Sheet of Ageing PhenomenaB: Ageing management program every 10 years

C: Technical evaluation of ageing phenomenaD: Implementation of seismic safety evaluationE: List of ageing phenomenaF: Concept of ageing management

Example of screening of ageing

phenomena for actual equipment

using the Summary Sheet of

Ageing Phenomena

摩耗 ○

疲労割れ（フレッティング疲労割れ） ○

疲労割れ（高サイクル疲労割れ） ○

羽根車ステンレス鋼鋳鋼腐食（キャビテーション） ○

羽根車リング－摩耗 -

ケーシングリング－（消耗品・定期取替品）取替品

軸受箱鋳鉄，炭素鋼鋳鋼腐食（全面腐食） ○

軸受（すべり）－（消耗品・定期取替品） -

軸受（ころがり）－（消耗品・定期取替品）取替品

軸継手低合金鋼，炭素鋼摩耗ギア型でない

・

・

・

・

・

・

・

・

・

・

・

・

低合金鋼（内面ステンレス盛金）（想定されず） -

疲労割れ ○

応力腐食割れ ○


疲労割れ ○


ケーシングボルトクロムモリブデン鋼，低合金鋼腐食 ○

メカニカルシール－（消耗品・定期取替品）取替品

ガスケット－（消耗品・定期取替品）取替品

Oリング－（消耗品・定期取替品）取替品

ケーシングドレン管ステンレス鋼応力腐食割れ ○

・

・

・

・

・

・

・

・

・

・

・

・

台板炭素鋼腐食（全面腐食） ○

取付ボルト炭素鋼腐食（全面腐食） ○

基礎ボルト炭素鋼腐食（全面腐食） ○

余熱除去ポンプ

ケーシングカバー

ステンレス鋼鋳鋼

ポンプの

容量-揚程確保

主軸ステンレス鋼

経年劣化事象

バウンダリの

維持

機器の支持

ケーシング


機能達成に

必要な項目

部位材料

摩耗 ○










・

・

・

・

・

・

・

・

・

・

・

・


疲労割れ ○



疲労割れ ○







・

・

・

・

・

・

・

・

・

・

・

・







ポンプの

容量-揚程確保


経年劣化事象

バウンダリの

維持

機器の支持

ケーシング


機能達成に

必要な項目

部位材料

摩耗 ○










・

・

・

・

・

・

・

・

・

・

・

・


疲労割れ ○



疲労割れ ○







・

・

・

・

・

・

・

・

・

・

・

・







ポンプの

容量-揚程確保


経年劣化事象

バウンダリの

維持

機器の支持

ケーシング


機能達成に

必要な項目

部位材料

Incorporates experience with maintenance management at actual plants and information from IAEA I-GALL

22

Part Material Ageing phenomena RHR pumpFunction

Amendments Added to “Code on Implementation and Review of

Nuclear Power Plant Ageing Management Programs 2015”

� Technical evaluation of plants in a long-term shutdown state was added.Adds provisions for the technical evaluation of degradation at a plant, which has

been in a long-term shutdown state due to an earthquake, accident or other

reasons.

� Tsunami resistance safety evaluations were added.

Provides for the requirements regarding the technical evaluation of ageing

phenomena for tsunami protection facilities

� Incorporation of the latest knowledge

・JSME: incorporated f environmental fatigue evaluation methods

・JEA: incorporated the guideline for verification of environmental qualification

of safety-related electric equipment and instrumentation NPPs

・Latest knowledge was incorporated in the “Guide for evaluation of ageing

degradation of cables at nuclear power plants”.

� Incorporation of information in IAEA I-GALL

Assumptions for the evaluation of wear, acid corrosion, etc. are added to the

Summary Sheet of Ageing Phenomena

� Provides for the methods for evaluation of seismic safety of pipes having

thinned walls

� Sorts out ageing phenomena for which seismic safety should be evaluated

23





Individual

equipment

（Example）

Reactor Vessel：

Core Internal：

Pipe：




Management

system

Human factor














PSR＋ Guideline


Syste

m/s

tru

ctu

re

/ co

mp

on

en

t

De

sig

n/

ma

nu

factu

ring

・Inspection

・evaluation

・repair

・Ageing

・monitoring/

・assessment



Development of

guidelines on

safety evaluation

is being studied

based on the

guidelines on

planning of

decommissioning.




24

Decommissioning

Plan/Implementation

Development of Codes/Standards to Address

Decommissioning of NPPs

Since the new regulatory requirements were put into force after the Fukushima Daiichi accident,

some NPPs have applied for the extension of operational periods, additional 6 NPPs have

decided on decommissioning. In response to demand for developing standards regarding the

decommissioning process, AESJ has developed/revised following standards:

� “Planning of Decommissioning for Commercial Power Reactors: 2011” AESJ-SC-A002:2011

Specifies the basic concept and technical requirements* for planning the decommissioning

of a commercial power reactor

* basic policy of planning, updating field conditions, management of nuclear fuel material,

planning of work, disposal of contaminated material, assurance of safety, etc.

� “Implementation of Decommissioning for Nuclear Facilities : 2014” AESJ-SC-A003:2014

Specifies the basic policy for implementation of decommissioning and the methods for

implementation

* basic policy of implementation, updating field conditions, management of nuclear fuel material,

work activities (safe storage, decontamination, dismantling, and maintenance), disposal of

contaminated material, assurance of safety during decommissioning, radiation control, etc.

� “Seismic Safety Guide for Nuclear Power Reactors in Decommissioning: 2013” AESJ-SC-

A006:2013

Specifies the basic policy for assurance of seismic safety during decommissioning, classification of

safety importance, method for confirming the seismic safety.

25

Current Further Activities

�Further development with taking into account the lessons

learned from the Fukushima Daiichi accident : (including

safety improvement measures against natural phenomena

and severe accidents)

�Establishment of a standardized process to ensure the safety

during decommissioning and standards for new generation

reactor plants

�Continuous update by introduction of new technologies,

incorporation of latest knowledge and R&D results

�Development / Improvement of codes/standards for PRA,

IRIDM (Integrated Risk-Informed Decision Making).

26

It is also necessary to correspond to

continuous safety enhancement evaluation activities

and new regulatory inspection system .

Safety Enhancement Evaluation (1/2)

Refer to https://www.nsr.go.jp/data/000195229.pdf

Evaluation for ensuring safety of Nuclear Power Plant

Outline of the System

<Objective>

• Utility has a responsibility to take necessary measures such as the setting of additional facilities and

safety culture to the nuclear safety improvement for while being based on the latest knowledge.

• It is intended to ensure and enhance the nuclear safety by the continuous improvement of facilities

(hardware) and safety culture (software) across a framework of regulation.

<Frequency of the report>

• Submit within 6 months after completion of Periodic Inspection of the Facilities

<Openness>

• Release after submission to NRA

Contents of report

27

① The range where the compatibility to laws and ordinances was approved by regulation

・Latest design based on contents such as application for permission for the establishment

・Latest applied measures based on the Operational Safety Programs

② The applied measures for voluntary safety enhancement

・Explanation about the safety enhancement

③The evaluation and analysis of applied measures for ensuring safety review

③-1 Evaluation of the activity to safety enhancement [every 5 years]

・Re-evaluation of internal & external events, PRA, evaluation of MS (margin of safety; stress test)

③-2 Medium-and-long term evaluation of the activity to safety enhancement [every 10 years]

・Plant design, the safety performances, reflection of the knowledge such as other plants, etc.

④ Comprehensive safety assessment

http://www.kyuden.co.jp/var/rev0/0079/6244/xd8kmzu5.pdf

Summary of the Safety Enhancement Evaluation Report on Sendai Unit 1

As for ensuring the safety of nuclear power plant in operation, it is operator’s own activities to improve and enhance safety, the reliability of the facilities continuously, and carry it out with the goal of reducing the risk of the nuclear power plant reasonably.

Checks implementation status of these maintenance activities in the operational safety inspection.

Safety assessment which is conducted to ensure appropriate safety margin is maintained.

Every 5 year except for a large scale designed change

Comprehensive safety assessment

Extraction of further safety enforcement and improvement measures・Hardware（countermeasure of SSC），・Software（Management）

Tutorial on Probabilistic Risk Assessment (PRA)

2017.7.6

28Safety Enhancement Evaluation (2/2)

New Regulation System for Nuclear Power Plant29

30

Major codes/standard to address the new inspection system (plan)

It is necessary to develop and revise related codes/standards so that the licensee can take necessary

actions to assume the primary responsibility for safety assurance, including quality control throughout

the plant life from design and construction through operation of nuclear facilities.

・JEAC4111 “Code on Management System for Nuclear Safety”

・JEAC4209/JEAG4210 “Code/Guideline for Maintenance Management at Nuclear Power Plants”

・Standards related to installation of modification of facilities to meet the new regulatory requirements

(facilities to deal with tsunamis, tornados, volcanic eruption, flooding, fire, severe accidents, etc.)

・JEAG4612 “Guideline for Importance Classification of Electric/Mechanical Equipment with Safety Functions”

・AESJ Standards related to PRA

・JEAC4212 “Code for Inspections of Core and Fuel at Nuclear Power Plants”

・”Rules on Design and Construction for Nuclear Power Plants”, “Rules on Fitness-for-Service for NPPs”

・Other standards

Development of Codes/Standardsaccording to New Regulation System

The new inspection system (integrated oversight/evaluation by the regulatory body: ROP etc.) will be

introduced in 2020, which specifies the utility’s responsibility to enhance and maintain nuclear safety and

to provide the total quality management system from design/construction stage to O&M stage.

The utility utilizes risk information and data showing the safety assurance level as objective indicators to

achieve effective maintenance activities according to the safety importance.

It is also necessary to enhance the transparency of safety assurance activities through the use of

consensus Codes/Standards and promote public understanding in safety assurance activities through

disclosure of information and communication.

AESJ (Atomic Energy Society of Japan) Challenges of Evolution for the Risk-Informed Decision Making Standard

• 2010: The SC (Standards Committee) of AESJ developed a standard for RIDM targeted for changes in safety related activities in NPP.

• 2011: The Fukushima Dai-ichi Accident has occurred.

Japanese Utilities should adopt risk-informed approach for their voluntary safety improvement activities.

- OLM, AOT, RI-ISI, SAM

- Safety measure against External Hazards

New regulations has been introduced.

- Safety Improvement Assessment (Publication of Safety Improvement measure)

- New Inspection System(Reactor Oversight Process)

Utilities Regulator

• The SC of AESJ is developing a New “Integrated” RIDM Standard.

- Optimized Decision: Integration of various key elements

- Accountability: A systematic approach

- User-friendly: Basic Requirements and practical approaches

R.G.1.174

Concept

INSAG-25

Concept

Refer to “Challenges of Evolution for the Risk-Informed Decision Making Standard”, Hayashi et al., ICONE25 July 6,2017

31

IRIDM process (AESJ Technical Report)• In 2016, the Standards Committee of the Atomic Energy Society of Japan

investigated and summarized an integrated decision making process utilizing risk information, AESJ technical report “The Concept of Adopting Continuous Safety Improvement Measures.”

• Now, SC is developing a new IRIDM standard to apply various risk applied work based on this technical report.

Refer to “The Risk-management for continuous safety improvement against beyond-design conditions of external events”, Narumiya, Post-SMiRT24, August 29,2017

32

� Developing and utilizing voluntary codes and standards are

important for safe and stable long-term operation of nuclear power

plants in future.

� The utilization of codes and standards, which have been established

under the principles of “fairness”, “equitableness” and “openness”,

can enhance the accountability toward members of the public.

In this regard, it is also necessary to promote the cooperative work

with overseas organizations to achieve the consistency with

international standards and contribute to the further improvement of

international standards.

� It is necessary to make a continuous effort to incorporate the

experience with operation and maintenance management, latest

technical knowledge and R&D results obtained in Japan and abroad

into the codes and standards so that they can be used in applying

advanced technologies to actual plants in an optimized manner.

Conclusion33

Thank you for your attention

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