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General Articles
Vol.1, No.4, GA10

Industry’s Efforts toward Technology Development Related to Aging Management of PWR Plants

PLM Group, Nuclear Power Division, The Kansai Electric Power Co.,Inc.

1. Foreword

The term, “aging plants”, which refers to nuclear power plants operating more than 30 years, has become popular among local residents around nuclear power plants as well as among the media. The term “aging” was first used in the title of the document issued by Agency of Natural Resources and Energy in 1996, “The Basic Concept of Aging Management”.

In addressing aging degradation, it was a matter of question which measure of time (date, month or year) should be used as a unit to consider an events. Although the aging plant was defined as a plant operating more than 30 years, the mode and timing of aging degradation depend on the environment and service conditions under which a component has been operating. Previous efforts made by related parties have contributed to the prevention of potential aging events which may challenge the reactor safety. Owing to such efforts, causes of some events attributable to defective design have been clarified and preventive maintenance measures, including the replacement and mitigation, have been taken accordingly. As a result, the number of events resulting from such aging phenomena has been reduced. On the other hand, the events caused by aging phenomena attributable to manufacturing processes (welding, surface treatment, etc.), which hardly emerge, are slightly increasing. There have been many cases that a shorter weld line due to lack of penetration in the narrow gap has been led to fatigue cracking. More recently, incidents of stress corrosion cracking due to work hardening and local tensile stresses on the surface have been observed.

It should be noted that an effort to analyze not only the phenomena but also the mechanisms of the above events to clarify the root causes can improve the quality of preventive maintenance by means of rolling-out the analytical results to relevant plants.
This paper introduces valuable experiences with the application of the results of technology development regarding to aging degradation of operating plants from the viewpoints of R&D and reactor operators.

2. Electric power companies' efforts to pursue aging-related R&D and apply the results to operating plants

2.1 History of aging-related R&D

The history of aging-related R&D, which has been systematically conducted since around 1985 and the organizational system of Japanese nuclear industry in developing the aging management strategy map 2007 are briefly described bellow.

2.1.1 History of Japan’s PWR (Fig.1)


2.1.2 Past Achievements and Future Prospects for Ageing-related Research Activities (Fig.2)


2.1.3 Outline of Next Generation LWR Development

Actions based on Nuclear Power Nation Plan
1)  Develop next generation reactors which can be recognized as the world standard
(one each type of 1,700~1,800MW-class *1 BWR and PWR)
  *1 : Options also include the 800~1,000MW-class reactors which do not interfere with the standardization.
2)  The government and electric power companies should jointly work on the technology development projects led by the manufacturers. Need to develop the next generation reactors intended not only for the domestic market but also for the international market.
Term of development: FY2008~2015 *2 (8 years)
  *2 : The material tests requiring a long period of time and some technology development projects will be continued subsequently.
Subjects of Development (6 core concepts)
- Achieve significant reduction of spent fuel and the world’s highest availability by developing the NSSS utilizing the world’s first 5% or higher enrichment fuel.
- Realize standardized plant design regardless of site conditions by adopting the seismic isolation technology.
- Integrate the development of new materials and improvement of water chemistry aiming at 80-years of plant design life and significant reduction of radiation exposure received during the maintenance work.
- Reduce the construction work period significantly by introducing the radial construction technologies.
- Achieve both the world’s highest safety and economic efficiency by optimized combination of passive and active systems.
- Adopt the world’s most advanced plant digital technology to improve the availability and safety simultaneously.
  (Reference material from September 18, 2007 15th nuclear subcommittee meeting of Electric Power Industry Task Force, Advisory Committee for Natural Resources and Energy)

2.1.4 Industrial Efforts to Promote PLM-related R&D Activities (Fig.3)


2.1.5 Studies Performed by 10 Sub Groups

Sub groups   R&D focus
  1. SCC
  2. Irradiation embrittlement
  3. Fatigues
  4. Seismic safety
  5. Concrete degradation
  6. Cable insulation degradation
  7. Pipe wall thinning
  8. Inspection/monitoring technologies
  9. Preventive maintenance & repair technologies

  1. Fundamental research projects
    Clarify ageing mechanisms, understand potential events, and develop alternative materials
  2. Development of advanced technologies
    Technologies required for evaluation, inspection and repair
  3. Application of technologies to actual units
    Technical approaches, demonstration tests, codifying process and approval by the government
  1. Enhancement of maintenance activities
  1. Construction of database
    Database required for PLM technical evaluations
  2. Improvement of management methods
    Development of alternative inspection methods replacing conventional overhaul inspections during plant shutdown
  3. Improvement of regulatory system
    Proposals for improvement of the current regulatory system
  4. Fostering and assurance of human resources
    Establish the system to pass on skills to the next generation and assure human resources

2.1.6 The History of PLM/AM at Japanese NPPs (Fig.4)


2.1.7 Replacement of Major Components at Mihama#3 (Fig.5)


Several examples of difficulties we have faced in applying the R&D results to operating plants are introduces in this article.

Following incidents of SCC in low pressure turbine rotors and discs at nuclear power plants in the USA and in Japan, partial repair and replacement of concerned sections had been conducted. Then, considering the results of R&D, it was determined to adopt the integrated rotor and disc designs to improve the SCC resistance. Similar to the surveillance tests for the reactor vessel structural materials, accelerated aging tests were conducted by placing test pieces made of same material as actual turbine parts into the casing of a steam turbine. From the test results, the residual life of turbine parts was evaluated.

The next subject of aging was irradiation assisted stress corrosion cracking (IASCC) in baffle former bolts. The research project on this phenomenon was initiated before any incident attributable to this phenomenon happened. Since the discovery of IASCC in the baffle former bolts at Bugey, France 15 years ago, research projects have been promoted and an international network has been established.

The last one is primary water stress corrosion cracking (PWSCC). Since the discovery of PWSCC in the rolled expansion zone of steam generator tube at a PWR plant, which was called “Achilles’ Heel”, incidents of PWSCC have been observed in the reactor vessel penetration nozzles at Bugey-3, in France where the inside diameter was expanded by machining, and then at Ohi-3 in the RV head nozzle welds, which had been subject to grinder polishing. These PWSCC were supposed to be caused by residual stresses on the component surface. Recently, PWSCC has been found in pipe nozzles of replaced steam generators. All these cracks occurred in nickel-based alloy components. However, it should be noted that in some cases, cracking in stainless steel material has propagated to an extended length. There is growing expectation for new findings to be obtained from the future R&D.

2.2.1 Application of Developed Technologies

1) Stress corrosion cracking/corrosion fatigue cracking in turbine discs
  • Considering incidents at NPPs in the US and Japan, local investigations were performed and then the results were incorporated in the total turbine maintenance. Data used to perform the investigations by simulating the actual unit were accumulated.
  • Corrosion fatigue tests were conducted using test pieces and steam taken from actual units.
2) Irradiation assisted stress corrosion cracking (creep, swelling) in baffle former bolts
  • Degradation of baffle former bolts can be predicted before experiencing any troubles.
  • Construction of an international network and 15-year history of IASCC Advisory Committee.
  • Research project regarding swelling, tests using specimens taken from a decommissioned plant in France (jointly by EDF-WH-INSS).
  • Inspection/repair technologies---Focus on actual results of application (technology transfer from a German company)
3) PWSCC in Ni-based alloy (Inconel 600)
  • Problem with SG tube---SCC in the tube expansion subjected to roll machining
  • Problem with RVH in France---SCC in enlarged nozzle ID subjected to drilling
  • Ohi-3 RVH cracks---SCC in welds subjected to grinding
    arrow2The common cause of SCC is cracking in the surface subjected to machining
  • Mihama-2 SG tube cracking---SCC in welds subjected to grinding
4) PWSCC in stainless steel component
  • Mihama-2 SG tube cracking---Propagation of cracks were observed only in the areas subjected to reinforcement machining

2.2.2 Inspection of Steam turbines (Fig.6)

- Low Pressure Turbines Replacement Work
Objective of replacement work
Considering incidents of stress corrosion cracking (SCC) in low pressure turbine discs observed at overseas NPPs, Kansai has decided to replace Ohi-3 low pressure turbine with partially integrated design rotor with one adopting less SCC sensitive integrated design rotor. At the same time, the high pressure turbine will be also replaced in order to assure long-term reliability. In replacing low/high pressure turbines, advanced technologies intended for the improvement reliability and efficiency will be applied.
Schematic drawing of new low pressure turbines (Fig.7)
EJAMGA1-4-Tanaka(Kepco)Fig.9_1(s) 1. Preventive maintenance against SCC EJAMGA1-4-Tanaka(Kepco)Fig.9_2
1) Adoption of integrated rotor
arrowintegral rotor made of low strength material with less SCC sensitivity. [690MPa -› 620MPa]
2) Adoption of large blade roots and grooves

2. Technologies for reliability EJAMGA1-4-Tanaka(Kepco)Fig.9_3improvement
3) Adoption of ISB
arrowISB design utilizing blade untwisting forces can reduce vibratory stresses

2.2.3 Case of deterioration : Baffle former bolts (Fig.8)

- Technical evaluation : Degradation modes of Core Internal(PWR) (Fig.9)
EJAMGA1-4-Tanaka(Kepco)Fig.11(s) ○ Baffle former bolts
IASCC of SUS347,316
All bolts of Mihama1 and
2 have been replaced

○ Barrel former bolts
IASCC of SUS347,316
Judging from the level of degradation of baffle former bolts, stress and temperature, barrel former bolts are OK in the foreseeable future.
- Technical Evaluation of IASCC for Ohi#1(1/2) (Fig.10)
- Technical Evaluation of IASCC for Ohi#1 (2/2) (Fig.11)
EJAMGA1-4-Tanaka(Kepco)Fig.11Evaluation of integrity

○ The result of baffle former bolt evaluation using JSME Fitness-for Service Code shows that the possibility of IASCC cannot be ruled out when the operation hours exceed about 250,000 (i.e., maximum radiation dose exceeds 6x1022 n/cm2 = about 40dpa).

○ However, the radiation dose after going through about 250,000 hours for the bolts, which are expected to be damaged in the early stage, is low (<20dpa). In addition, tests using the latest knowledge show a similar result. Therefore, there cannot be a significant change in the trends of IASCC.

Actions to address ageing phenomena

○ Study the future maintenance programs, including inspection and replacement, by utilizing JSME Code on Fitness-for-Service.

○ Actively participate in the national project “Development of evaluation technologies regarding IASCC”, technology development projects sponsored by private organizations and meetings to develop codes and standards, and consider the applicability of those results to the actual maintenance program.

- Integrity Evaluation of BFB - Multiple Ageing Phenomena - (Fig.12)

2.2.4 Main Components Using Alloy 600 in J-PWR (Fig13)

- The History of Trouble and PM of RV-Head (Fig.14)
- Case of deterioration : RV-haed penetration (1)(Fig.15)
- Case of deterioration : RV-head penetration (2)(Fig.16)
- Experiences in Japan
-1. Discovery of Cracking
[Kansai Electric’s Mihama-2]
During the periodic inspection at Mihama-2, ECT was conducted to confirm the integrity of SG nozzle welds before applying shot peening, which aimed at reducing residual stresses applied on SG nozzle welds. As a result, a significant indication was detected on A-SG inlet nozzle weld. Subsequent UT confirmed the crack depth which caused the remaining wall depth below the technical criteria. (Fig.17)
-2. Results of Inspections at Other NPPs
At the other Japan’s NPPs, ECT of 600 series Ni-based alloy weld ID of SG inlet nozzles was performed. As a result, significant indications were found at some plants and at those plants UT was performed for the purpose of depth sizing. For several cracks, their depth could be measured by UT. The table below summarizes the inspection results.
Plant Operating hours Loops Indications by ECT UT results (length shows ECT results)
Kansai Mihama - 2* About 93,000 2 A-inlet nozzle weld: 13 Max. length : about 17mm
Max. depth : about 13mm
JAPC Tsuruga - 2** About 151,000 4 A-inlet nozzle weld: 1

Detected (the depth and length sizing was impossible although a significant indication was confirmed.)
B-inlet nozzle weld: 5 Max. length : about 21mm, Max. depth : about 12mm (an another crack having a similar size was also confirmed)
C-inlet nozzle weld: 23 Max. length : about 14mm, Max. depth : about 13mm (cracks were also found at other 6 locations)
Kansai Takahama-2 About 100,000 3 A-inlet nozzle weld:3 Not detected
B-inlet nozzle weld:2 Max. length: about 7mm, Max. depth: about 6mm
C-inlet nozzle weld:4 Max. length: about 14mm, Max. depth: about 8mm
Kansai Takahama-3 About 175,000 3 A-inlet nozzle weld:7
Max. length : about 28mm, Max. depth : about 9mm
B-inlet nozzle weld:16 Max. length : about 38mm, Max. depth : about 15mm
C-inlet nozzle weld:9 Max. length : about 12mm, Max. depth : about 9mm
Kyushu Genkai-1 About 96,000 2 A-inlet nozzle weld:3 Not detected
Hokkaido Tomari-2 About 131,000 2 A-inlet nozzle weld:3 Max. length : about 13mm, Max. depth : about 7mm
B-inlet nozzle weld:10 Max. length : about 10mm, Max. depth : about 5mm
->Detailed investigation was performed at the representative plants
* Investigation by cutting the pipe section
** Tests using boat samples
-3. Detailed Investigation (Fig.18)
SUMP examination of the section in A-SG inlet nozzle weld where the deepest crack (about 17mm long and 13mm deep) was confirmed showed that several axial cracks having length of about 3 to 5 mm extended intermittently and along the interdendritic boundaries of the 600 series Ni-based alloy weld. The crack had similar characteristics to PWSCC which had been observed in 600 series Ni-based alloy welds of NPPs both at home and abroad.
-4. Suspected Causes (Fig.19)
1) SG nozzle weld ID was subject to machining: Mihama-2, Takahama-2, Genkai-1
The welding and machining processes applied to the replacement SGs during the manufacturing phase resulted in high residual tensile stresses on the SG inlet nozzle weld ID, which caused the initiation of PWSCC and the PWSCC propagated along interdendritic boundaries in the axial direction.
2) SG nozzle weld ID was subject to grinder finishing and buffing while repair welds were subject to grinder finishing: Tsuruga-2, Takahama-3, Tomari-2
It is suspected that in the SG nozzle welds were subject to grinder finishing and buffing and partially subject to grinder finishing after repair welding, high residual tensile stresses were generated in the areas with traces of grinder finishing, which caused the initiation of PWSCC, and the PWSCC propagated along interdendritic boundaries in the axial direction.
-5. Corrective Actions to Addresss PWSCC
Regarding the concerned weld and safe end, which were cut out for the cause investigation, the safe end was replaced with a new one and a corrosion resistant 690 series Ni-based alloy weld was applied between the inlet pipe and safe end. For the assurance of safety, buffing was applied to reduce residual stresses.
Tsuruga-2, Takahama-2, Takahama-3, Tomari-2
After removing shallow defects by cutting the entire circumference of the concerned weld ID, deep cracks were removed by grinding and then corrosion resistant690 series Ni-based alloy weld overlay was applied to the entire circumference of ID. For the assurance of safety, buffing was applied to reduce residual stresses.(Fig.20)
After grinding the concerned area to remove defects, ultrasonic shot peening (USP) was applied to reduce residual stresses.

2.2.5 Non-Destructive Testing Technique
- Stream Generator Nozzle Safe End Weld - (Fig.21)


2.2.6 Preventive Maintenace (Water Jet Peening) (Fig.22)


2.2.7 Material Improvement - Alloy 690 cladding

Alloy 690(52) cladding for Reactor Vessel nozzle safe-end (Fig.23)

2.2.8 PWSCC in Stainless Steel Components
- Fukui Prefecture Ageing Management Investigation Committee in FY 2006

Background and Achievements
EJAMGA1-4-Tanaka(Kepco)Fig.24 (Background)
At BWRs, the initiation and development of SCC have been reported even in stainless steel components, such as the core shroud and recirculation pipe, which are subjected to hard cold work (machining in room temperature). On the other hand, under the PWR environment, no SCC in stainless steel components has been observed. (Fig.24)
A verification program regarding SCC under the PWR environment should be included in the ageing management activities.
(Achievements obtained so far)
- Laboratory experiments revealed that cracks caused by SCC would develop even under the PWR environment if cold work was added.
- The crack growth rate in PWR increases as the yield stress (degree of cold work) increases like BWR.
Issues to be addressed
1. Clarification of SCC mechanism in cold worked stainless steel parts
2. Identification of conditions simulating an actual unit and data sampling for tests and evaluation
3. Investigation of SCC prevention measures
Aging management is intended to develop appropriate maintenance measures and schedules, and implement them according to the predetermined service period. However, the idea that the replacement of a component at any timing is the best possible measure may cause risks because the degree of aging depends on the manufacturing and work conditions in some cases. Therefore, it should be noted that the projection of the timing of aging provides only reference information. In this regard, it is necessary to implement plant operation and take steady steps for risk management by referring to the projected timing of aging while understanding that there remain R&D issues to be resolved.

3.1 Application of Lessons Learned from Past Events/Failures (Fig.25)


3.2 Results of Ageing Management Study and Total Management System (Fig.26)


3.3 Preparedness for the maintenance activities (viewpoint of information) (Fig.27)


3.4 Conclusion

1) It is necessary to fully understand the actual status in the field if insights cannot be obtained from laboratory tests.

2) Confirmation of the consistency between the current knowledge and actual status versus time is the key to clarifying the cause of an event.
If cause investigation has not be fully conducted, a similar event occurs in the field.
It is necessary to clarify the degradation mechanism in order to complete the cause investigation.

3) For the purpose of the Plant Life Management, maintenance activities are conducted based on an assumed operation period (timing of rebuilding the plant). Therefore, it is likely to presume that future trouble can be prevented by replacing the components which are expected to experience trouble in the near future considering the operation records achieved in a half of the assumed plant life.
However, the idea of adopting the unconditional replacement may be hazardous because the replacement work conditions may affect on degradation. Therefore, thorough consideration is needed before determining the actions.