0. Plant Life Management in Japanese PWRs
Kazuo Sakai
WANO Tokyo Centre

1. Contents Regulatory Position and Utilities’ Action
Outline of Technical Evaluation
Maintenance Policies Based on Technical Evaluation

2. Operating Years before the First PLM Report Publication
(Unit)
(March 1999) 15
Total : 51 units
PWR: 23 units
BWR: 28 units 13 11 9 10 7 7 4 4 5
・This graph shows the operating years distribution of light water reactor plants operating in Japan.
・Since the first commercial plant started operation in 1970, 10 utilities have built and operated 51 commercial LWRs.
・We here at Kansai have 11 PWR plants.
・Although the current operating condition is good, the number of plants having operated for a long period of time is increasing.
0-4
5-9
10-14
15-19
20-24
25-29
Operating Years

3. The First PLM Demonstration
1994-
1996
1998
2000
Technical Evaluation of Long-term Integrity
Tsuruga-1 , Mihama-1 , Fukushima 1-1
PART 1 (METI)
PART 2(Utilities)
・Major component
・Assumption:
60 years operation
Conclusion:
60 years operation is possible with proper long-term maintenance program
All components of plants
Assumption: 60 years operation
Conclusion:
60 years operation is possible with proper long-term maintenance program.
Proposal of long-term
maintenance program
Application to All Nuclear Power Plants
・First, I would like to explain the process of total plant life management activities.
・The evaluation consists of two steps:
・Part 1 of the evaluation is based on the basic policy on aged nuclear power plants provided by the Ministry of International Trade and Industry in April, 1996.
・Utilities voluntarily conducted more detailed technical evaluations on measures against aging phenomena based on the review by MITI (Part 1).
・These voluntary technical evaluations by the utilities are called Part 2.
METI:　Ministry of Economy, Trade and Industry

4. Maintenance Activities at a NPP
General Maintenance Cycle
Activities for Improvement
Understanding/evaluation of aging phenomena
Periodic Safety Review
(every 10 years）
Comprehensive evaluation of operating experiences
In operation
Annual inspection
Verification of integrity
Operation monitoring
Inspection
Incorporation of latest technical information
Walk downs
Probabilistic safety assessment
Repair/
replacement
Feedback to inspection plans
Plant Life
　　　　　　Management
Surveillance tests
(30 years after start of operation)
Modification
Condition monitoring
Incorporation of trouble experiences at domestic and overseas plants
Feedback

5. Cyclic PLM Process
Establish the long-term maintenance program in the first PLM review, before 30th anniversary
Implement the long-term maintenance program as planned during annual outages.
Conduct a next PLM review during the PSR performed every 10 years
Periodic Safety Review
(every 10 years)
30th anniversary ▼ ▼ ▼
PLM Review
Annual
outage
Annual
outage
PLM Review
Annual
outage
Long-term maintenance program
Long-term
maintenance program

6. PLM Report Submitted Plant
Japan Atomic Power
Tsuruga 1 (BWR,357MW,1970)
Kansai EPC
Mihama 1 (PWR, 340MW,1970)
Mihama 2 (PWR, 500MW,1972)
Takahama 1 (PWR, 826MW,1974)
Takahama 2 (PWR, 826MW,1975)
Kyushu EPC
Genkai 1 (PWR, 559MW,1975)
Tokyo EPC
Fukushima Daiichi 1 (BWR, 460MW,1971)
Fukushima Daiichi 2 (BWR, 784MW,1974)
Chugoku EPC
Shimanne 1 (BWR, 460MW,1974)
Note: (Reactor Type, Electric Capacity, When Commercial Operation started)
・This table lists our 11 PWR plants.
・As illustrated in this table, Mihama unit no. 1 has 29 years operational experience at present. For the subsequent plants, 7 plants will have been operating for more than 20 years.
・Therefore, safety and reliability assurance at aged plants becomes an important issue.
・At these aged plants, replacement of major components including the steam generator and the reactor vessel head has been performed. The steam generators with MA Inconel 600 tubing have been completely replaced. Vessel head replacement work has been and presently being performed at our plants.

7. Technical Evaluation Process of PLM
Select the components related to the plant safety and reliability, and select the representative components to be evaluated considering type, operating environment and material.
Identify possible ageing phenomena, and identify parts of the components and their sections where aging phenomena will occur in the future, considering environmental condition (temperature, pressure, water chemistry), material, design bases, research results and operation experiences.　
Evaluate the ageing phenomena of the identified parts and effectiveness of present maintenance programs.
Establish maintenance policies and long-term maintenance programs of the components, and identify R&D subjects such as the improvements of inspection technique, repair technique, mitigation technique, replacement technique, and material degradation mechanism.

8. Components to be Evaluated
Pump
Heat exchanger
Pump motor
Vessel
Piping
Valve
Reactor internals
Electrical cable
Electrical equipment
Turbine &
associated equipment
Concrete structures
Instrumentation &
control system
Ventilation system
Mechanical equipment
Power supply system
Other systems
・The utilities evaluated all plant systems, structures and components.
(Approx. 30,000 components in total)

9. Technical evaluation Result
Most of NPP components are repairable/replaceable.
The components will be able to be operated with proper maintenance programs on general maintenance cycle (inspection, repair and replacement).
(Approx. 30,000 components in total)
Approximately 10 components are difficult to be replaced.
Next Slide

10. Policies for Components Difficult to be Replaced
Newly developed technologies have enabled replacement of the components, such as steam generators and R/V heads.
However, integrated replacement of such components requires considerable time and costs. Therefore, it is necessary to predict aging phenomena for each part and take preventive actions in advance.
With proper maintenance programs, we evaluate our plants will be operated more than 60 years.

11. Components Difficult to be Replaced
Containment Vessel
Pressurizer
Reactor Vessel
Steam Generator
Concrete Structures
Electrical Cable
Main Coolant Pipe
ＰＡＲＴ１は、これらに示すように、補修及び取替が容易でなく、かつ安全上重要な機器・構築物として、８つの機器と１つの構築物をＰＬＭの検討を行う上での評価対象とされました。
・原子炉格納容器　　　　　　　　　　　　・加圧器
・原子炉容器　　　　　　　　　　　　　　・蒸気発生器
・コンクリート　　　　　　　　　　　　　・ケーブル
・１次冷却材管　　　　　　　　　　　　　・１次冷却材ポンプ
・炉内構造物
Reactor Internals
Reactor Coolant Pump

12. Aging Phenomena of Reactor Vessel
Next Page
・This is an example of aging phenomena assumed for the reactor vessel at Mihama unit No.2.

13. Aging Phenomena of Reactor Vessel
・The reactor vessel is divided into parts, and aging phenomena including wear, corrosion, fatigue cracks, thermal aging and material degradation are identified for each of these parts. These aging phenomena were evaluated considering the boundary maintenance as a required function.
・According to the evaluation, the ones that are circled are aging phenomena which are considered to be important for measures utilized against aging phenomena; the ones marked with triangles are not considered to be significant from an engineering point of view.

14. Example of Technical Evaluation of 10 components(1/2)
機器毎に想定される経年変化事象とその評価結果をまとめたものがこの表です。
評価として、健全性評価結果と高経年化への対応という形でまとめることができます。
原子炉容器でいいますと、経年変化事象として抽出した疲労に対し、疲労解析により、疲れ累積係数が１以下であることを確認し、６０年間の運転は可能であるという結果が得られましたが、高経年化への対応として、疲労解析の値は実過渡回数に依存することから、定期的にその回数確認が必要であるという留意事項が挙げられております。
また、下部胴の中性子照射脆化につきましても、中性子照射脆化予測式に基づき、評価上６０年間の運転は可能でありますが、今後も適切に監視試験片の取り出しを行い、評価結果の妥当性を確認する必要があるとされております。
インコネル６００合金使用部位のＳＣＣについては、上部ふた部は、取り替えが計画されており、その他の部位は、計画的な検査を行う必要があるとされています。

15. Example of Technical Evaluation of 10 Components(2/2)
機器毎に想定される経年変化事象とその評価結果をまとめたものがこの表です。
評価として、健全性評価結果と高経年化への対応という形でまとめることができます。
原子炉容器でいいますと、経年変化事象として抽出した疲労に対し、疲労解析により、疲れ累積係数が１以下であることを確認し、６０年間の運転は可能であるという結果が得られましたが、高経年化への対応として、疲労解析の値は実過渡回数に依存することから、定期的にその回数確認が必要であるという留意事項が挙げられております。
また、下部胴の中性子照射脆化につきましても、中性子照射脆化予測式に基づき、評価上６０年間の運転は可能でありますが、今後も適切に監視試験片の取り出しを行い、評価結果の妥当性を確認する必要があるとされております。
インコネル６００合金使用部位のＳＣＣについては、上部ふた部は、取り替えが計画されており、その他の部位は、計画的な検査を行う必要があるとされています。

16. Maintenance Policies against Aging Phenomena (1/４)
Research and development as necessary
Part
Section
Maintenance policies
R/V head penetration nozzle
Base metal and weld
Integrity verification by inspections
Timely replacement of R/V head
Inspection technologies
Lifetime evaluation technologies
PWSCC
（Immediate　actions）
Inspection and depth sizing technologies
Lifetime evaluation technologies
Repair technologies
Technologies to evaluate crack propagation and allowable limits
Integrity verification by inspections
Replacement as necessary
R/V hot/cold leg nozzle-to-pipe weld
Weld
PWSCC
（Future actions）
Operating continuously based on the result of evaluations of crack propagation

17. Maintenance Policies against Aging Phenomena (2/4)
Research and development as necessary
Part
Maintenance policies
Section
Inspection and depth sizing technologies
Water jet peening technology
Lifetime evaluation technologies
Repair technologies
Replacement technologies
Technologies to evaluate crack propagation and allowable limits
(Immediate　actions)
R/V bottom-mounted instrumentation nozzle
Base metal and weld
PWSCC
Integrity verification by inspections
Prevention of SCC
Repair as necessary
(Future actions)
Operating continuously based on the evaluations of crack propagation

18. Maintenance Policies against Aging Phenomena (３/４)
Research and development as necessary
Part
Section
Maintenance policies
（Immediate actions）
Integrity verification by inspections
Replacement of bolts
（Future actions）
Operating continuously based on the acceptance criteria for bolt failure
Inspection technology
Bolt replacement technology at a higher speed
Material with high IASCC resistance
Technology to estimate possibilities of IASCC
C/I baffle former bolt
Bolt neck
IASCC

19. Maintenance Policies against Aging Phenomena (４/４)
Research and development as necessary
Part
Section
Maintenance policies
Primay coolant pipe
Base metal and weld
Thermal aging and fatigue crack
Integrity verification by inspections
Integrity evaluation by elastic-plastic fracture mechanics evaluation
Integrated replacement if integrity can not be assured by means of above evaluation
Inspection and defect sizing technology
Optimized repair method
Optimized replacement method
Technology to estimate decrease in toughness
Elastic-plastic fracture mechanics evaluation method

20. Example of Predicted Aging Phenomena
in Reactor Vessel
Hot leg nozzle-to-pipe weld
SCC in alloy 182
Bottom-mounted
instrumentation nozzle
SCC in alloy 600/182
R/V head penetration nozzle
　　SCC in alloy 600/182
Alloy 182 is equivalent
to alloy 600
Lower shell
Irradiation embrittlement

21. Core Internals
Buffle Former Bolt
Bolt Crack (SCC)
US, France Plant

22. Fatigue Cracking in R/V Safety Injection Nozzle
Integrity evaluation: fatigue evaluations based on the past operating records
　 　The cumulative fatigue factor is sufficiently below the allowable limit
Allowable limit
（Change of temperature）
1.0
（Change of pressure）
Design cumulative fatigue factor
　　　 　（0.193）　×
Cold water
Cumulative fatigue factor expected during
the design phase
R/V
Safety injection nozzle
　Cumulative fatigue factor at the time of PLM evaluations
　（approx or less）　　×
Fatigue will be accumulated if changes of temperature and stress during startup and shut down as well as entrance of cold water during an accident are assumed.
30 years
60 years
Operating years
Maintenance measures for the aging plants
　Continued evaluations based on the number of actual transients
Avoid fatigue cumulative operations

23. The parts of Alloy 600（Reactor Coolant Pressure Boundary）

24. The Parts Made of Alloy 600 in S/G
Tubes
　Alloy 600
Safe-end welds of reactor coolant Inlet/outlet nozzles
Alloy 600

25. The Parts Made of Alloy 600 in Pressurizer
Safe-end welds of spray line nozzle (Alloy 600)
Safe-end welds of relief valve nozzle/ safety valve nozzle (Alloy 600)
The Type of Nozzle Structure
Type
Nozzle Structure
Safe-end (Stainless Steel)
Stainless Steel Welds
Nozzle (Low Alloy) A
Lining sleeve (Stainless Steel)
Safe-end (Stainless Steel)
Alloy 600 Welds
Nozzle (Low Alloy) B
Lining sleeve (Stainless Steel)
Safe-end (Stainless Steel)
Nozzle (Low Alloy)
Alloy 600 Welds C
Stainless Steel Clad
Safe-end welds of surge nozzle (Alloy 600)

26. SCC in R/V Head Penetration Nozzle
Base metal
Weld metal
Maintenance measures for the aging plants
UT inspections are performed during annual outages.
R/V head nozzle weld was replaced with Inconel 690 weld with corrosion resistance.
Replacement of material
Collect information
about operational experience
Alloy　 　　Alloy
　６００　　　　　６９０
（Improve SCC resistance）

27. Water Jet Peening （ＷＪＰ） for BMI
High pressure water injection in water generates cavitation bubbles by water jet and they collapse on the metal surface. Then the impact pressure changes tensile stress to compressive stress on the metal surface.
Use of WJP
Image of WJP
BMI
R/V
Water
Nozzle
Jet with bubble
Tip Nozzle for WJP
(Rotating motion ＋vertical motion)
Relevant Part
WJP Effect
Residual stress：-400MPa（surface）
　　　　 mm（Effective depth）
BMI

28. Replacement Method of BMI
Lifting of various tools
In the water
In the air
Remote control stage
Guide tube
Upper core internals
Lower core internals
Example: Removal of old BMI by cutting
Water seal cylinder
Cutting device

29. Repair Method from Outside of R/V
Existing alloy clad welding is used.
BMI
Alloy 600 welding
R/V (low-alloy steel)
Alloy 600 clad welding (existing)
IASCCについては、実験データ等が少なく、知見が少ないのですが、
その中から定数を検討しました。
Ｎは、定荷重 Ｑ ｍ
Repair by cap

30. Repair Method of BMI Replacement to the one with the same structure
In practical, may be expensive
Repair to maintain the same function
May be implemented based on crack depth
Example
Small cracks inside the tube
Removal of cracks by drill
先ほどまではどこまで損傷が許容されるかという話でしたが、ここでは損傷が
どのように進むかを検討しました。
Ｎｉ基合金のＳＣＣ予測式を用いて、各ボルトについての損傷時間から
ボルト損傷本数の増え方を予測します。
ここで、応力、温度、照射量の各影響に対する定数ｎ，Ｑ，ｍが未知で
あるため、これを定める必要があります。
If enough strength is attained, we can take an alternative.

31. Cracking in R/V Hot Leg Nozzle-to-pipe Weld
in alloy 182 weld
Hot leg nozzle
(low-alloy steel)
Primary coolant pipe
(stainless steel)
Reactor vessel
<
V.C.Summer>
Ringhals-3,4 (2000)
V.C.Summer (2000) ＜
Cause of cracking ＞
SCC due to high residual stress
on alloy 182 weld caused by
repeated welding processes for
repair

32. Repair method for R/V Hot Leg Nozzle-to-pipe Weld
Primary piping(hot leg)
Concrete plug
R/V
Heat insulator
R/V support
Normal concrete
Cutting of spool piece
Removal of spool piece
Insertion&welding of new spool piece

33. Conclusion(1/2) PLM is conducted for aging plants operated 30 years,
and repeated every 10 years in Japan.
Three utilities submitted their PLM-Reports of five plants to the METI. METI approved the results of the evaluations upon review process, and did the Nuclear Safety Commission the result of METI’s review.
The PLM-Reports said that the plants can be operated for 60 years with proper maintenance programs.

34. Conclution(2/2)
Approximately 10 components are difficult to be replaced. Therefore, it is necessary to predict aging phenomena for each part, establish the maintenance policy including proper maintenance and R&D programs such as inspection, evaluation, mitigation, repair and replacement, and take preventive actions in advance.
Trouble experiences and R&D information should be continuously collected worldwide, incorporating them into maintenance programs. Therefore, international cooperation should be enhanced.