Plant Life Management in Japanese PWRs Kazuo Sakai WANO Tokyo Centre

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

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

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

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

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

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.

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.

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)

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

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.

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

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

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.

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。

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。

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

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

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

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

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

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

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. 0.001 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

The parts of Alloy 600（Reactor Coolant Pressure Boundary）

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

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)

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）

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） 　　　　 0.5mm（Effective depth） BMI

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

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

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.

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

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

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.

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.