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Plant life management activities of LWR plants in Japan
Nuclear Engineering and Design 207 (2001) 241– 248 www.elsevier.com/locate/nucengdes
Technical note
Plant life management activities of LWR plants in Japan Akiyoshi Minematsu * Department of Nuclear Power Engineering, Engineering R&D Di6ision, Tokyo Electric Power Company, 1 -1 -3 Uchisaiwai-cho, Chiyoda-ku, Tokyo 100 -0011, Japan Received 22 February 2000; accepted 30 August 2000
Abstract Plant life management activities of Japanese LWR plants have been conducted since the early 1990s by the utilities and MITI (Ministry of International Trade and Industry) cooperatively. In Japan, where the regulatory practices are different from those in the US, there is neither law nor regulation that prescribes a licensed plant life for nuclear power plants. When an annual inspection is completed without any problem, the next cycle of operation would be permitted and this cycle can be repeated. However, it is generally known that mechanical components and structures deteriorate as they get older. So, we consider it very important to evaluate the long-term integrity of major systems, structures and components of old nuclear power plants. Japanese plant life management study consists of two parts. Both parts of the study were carried out confirming the integrity for the long-term operation of the three oldest Japanese LWR plants: Tsuruga Power Station Unit No.1 (BWR), Mihama Power Station Unit No.1 (PWR) and Fukushima Dai-ichi Nuclear Power Station Unit No.1 (BWR). The Part 1 study was conducted for the purpose of obtaining an outlook for long-term safety operation and was completed in 1996. The Part 2 study was conducted ensuring the plant integrity for the long-term operation in terms of, not only safety, but also reliability. The results of the Part 2 study were made public in February, 1999. Then, the recommended maintenance items were to be added to the existing maintenance programs of the three LWR plants. © 2001 Published by Elsevier Science B.V. Keywords: Plant life management; Power plants; Power stations
1. Plant life management activities
1.1. Introduction
Presented at the 26th Water Reactor Safety Meeting, Bethesda, Maryland, USA, October 26-28 1998. * Tel.: +81-3-42164901; fax: + 81-3-35968544. E-mail address: [email protected] (A. Minematsu).
Already more than 29 years have passed since the first Japanese nuclear power plant went into operation. In Japan, nuclear power plants, both new and old have been operated and maintained in a similar way.
0029-5493/01/$ - see front matter © 2001 Published by Elsevier Science B.V. PII: S 0 0 2 9 - 5 4 9 3 ( 0 0 ) 0 0 3 8 6 - 1
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A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
In the early 1990s, as Europe and the US were establishing their technical studies on the plant life management and the plant life extension of nuclear power plants, Japan has also come to recognize the importance of plant life management activities. This paper briefly describes the current status of the activities of the plant life management activities of LWR plants in Japan and outlines the future plans.
1.2. Basic concept The Part 1 study was designed to work out the concept of the plant life management through evaluation to confirm the integrity of the major components during the long-term operation of nuclear power plants. In the actual evaluation, 60-year operation was assumed. For the Part 1 study, the three oldest LWR plants described above were selected as the representative plants. Seven or nine components that were the most important for safety plant operation and difficult to repair or replace were selected for the Part 1 technical evaluation. Those were, reactor pressure vessel, reactor internals, primary containment vessel etc. For the Part 2 study, we reviewed the wider range of components (amounting to thousands of items) not only from the safety point of view, but also from the perspective of avoiding an unplanned shutdown for the three oldest LWR plants. This part of the study was intended to develop measures against aging degradation and establish the evaluation technique for the plant integrity during the long-term operation.
60-year operation properly. For example, in the case of BWR plants, it was recognized that the planned measurement against the wall thickness of thinned carbon steel was necessary. Furthermore, it is necessary to have the scheduled inspections of the reactor internals whose material is not low carbon stainless steel to avoid damage due to IGSCC. In the case of PWR plants, it was recognized to be necessary that measures would be taken against stress corrosion cracking of the parts made of alloy 600, such as the steam generator tubes. However, it was confirmed that the plant integrity could be preserved basically by the continuity of the present maintenance methods.
2. Part 2 evaluation method
2.1. Procedure of the Part 2 e6aluation The evaluation methods used in the Part 2 study are as follows (the outline of the evaluation methods is as illustrated in Fig. 1).
2.1.1. Selection of components for e6aluation The purpose of the Part 2 study was to evaluate the long-term plant integrity in terms of safety and reliability. In accordance with this purpose, the components to be evaluated were selected and were classified into three categories as follows. 2.1.1.1. Safety related components. This category includes the components falling under PS-1, PS-2, MS-1 and MS-2 categories in the guideline2:
1.3. Results of the Part 1 study In the Part 1 study, the integrity evaluations were conducted for seven components (BWR case), those were reactor pressure vessel, reactor internals, primary containment vessel, primary coolant piping, primary loop recirculation pump, cable and concrete structure. Through the Part 1 technical evaluation, additional inspections or monitoring were recommended for some of the components to ensure the
2 ‘The Guidelines for Classification of Important Components for Light Water Reactor Facilities’, published by the Japanese Nuclear Safety Commission (NSC), provides the following definitions: PS, abnormality prevention system: loss of the function of the PS systems may subject the reactor facilities into the abnormal condition and thereby cause excessive radiation exposure to the general public or personnel; MS, abnormality mitigation (or suppression) system. The spread of abnormality in the reactor facilities will be prevented or put under control by the MS systems promptly, thereby preventing or suppressing the excessive radiation exposure to the general public or personnel.
A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
In the case of BWR — reactor pressure vessel, high pressure core injection pump, main steam isolation valve, feedwater piping, emergency diesel generator, air conditioning system of the central control room, etc. In the case of PWR — reactor pressure vessel, heat removal pump, pressure relief valve of pressurizer, auxiliary feedwater piping, emergency diesel generator, sea water pump, etc.
2.1.1.2. Reliability related components. This category includes components related to avoiding an unplanned shutdown, that fall under the PS-3 category in the guideline1: In the case of BWR — turbine, generator, generator excitation system, condensate system (including main condenser), feedwater system, circulating water system, main transformer, switching yard, etc. In the case of PWR — turbine, generator, generator excitation system, condensate system (including main condenser), extraction steam piping, main transformer, switching yard, etc.
243
2.1.1.3. Other important components. This category means all components which belong to the guideline except the safety and reliability related components. In the case of other important components, there are not as many differences between BWR and PWR. This category includes, fire protection system, radwaste disposal system, house boiler system, sea water intake system, etc.
2.1.2. Classification and selection of representati6e components In the case of BWR, the components to be evaluated amount to : 3000 items, according to the above selection criteria. In order to conduct the evaluation rationally, the grouping method was introduced. First, the selected components, as noted previously, were sorted into the component categories like pumps, motors etc.
Fig. 1. PLM evaluation procedure in Part 2 study.
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A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
Secondly, the components were further divided into groups according to grouping criteria, such as type, structure, operation circumstances (which include operated location, fluid properties, etc.) and materials. Thirdly, the representative components were selected from each group based on their importance, operating conditions (pressure, temperature) and other criteria.
2.1.3. E6aluation The representative components selected at Section 2.1.2 were broken down to the parts level and technically evaluated by considering the aging degradation phenomena. The evaluations for the representative components were conducted based on the experimental data, the actual plant operational experience etc. Combining these evaluation results and the current inspection and maintenance program of the components, the synthetic evaluation was carried out for the long-term operation. In case the current inspection and maintenance program of the component are judged to be insufficient, additional inspection and maintenance items should be recommended. The evaluation results for the representative components were applied to the other components which belong to the same group, considering the differences between the components. 2.1.4. Selection process of the aging degradation phenomena to be considered There are a lot of aging degradation phenomena in nuclear power plants. It is impractical to consider every possible phenomenon for each part. Therefore, the screening process was adopted to select those phenomena in terms of the environmental condition, the materials etc. The aging degradation phenomena that should be considered in this evaluation were selected as follows (see Fig. 2). 2.1.4.1. Compiling phenomena which occur in industrial materials and products. The aging degradation phenomena that were commonly found to occur in industrial materials and products (both electrical and mechanical) were selected. These phenomena were identified based on the survey of
literature, the study of past abnormalities and the latest technological information.
2.1.4.2. First phase screening of phenomena with respect to nuclear power plant en6ironment. The aging degradation phenomena that do not occur under the condition of light water reactors could be excluded. For example, it is not necessary to consider the halogenated corrosion (necessary to consider in the case of chemical plants) and the sigma phase embrittlement (necessary to consider in the case of temperature between 565 and 930°C). 2.1.4.3. Second phase screening of phenomena with respect to material en6ironment. The aging degradation phenomena that do not occur under the material environment could be excluded. For example, the carbon steel corrosion does not need to be considered when it is used in the anti-corrosive water quality environment. Also, it is not necessary to consider the thermal aging embrittlement for casting stainless steel used at low temperature. 2.1.4.4. Third phase screening of phenomena with respect to the experience. The aging degradation phenomena that can be justified technically to occur infrequently or with very little possibilities under the operating condition based on material test data, could be excluded. In the screening process, operational records and special factors that are specific to the plants should be taken into account. For example, it is not necessary to consider the inter-granular stress corrosion cracking of the austenitic stainless steel (304 stainless steel) under the environment of low concentration of dissolved oxygen (below 8 ppm) and low temperature (B 100°C). 2.2. Seismic safety e6aluation considering aging degradation 2.2.1. Purpose of e6aluation It is important to ensure whether the aged components remain sound or not under seismic loads.
Fig. 2. Principles of identifying possible aging degradation phenomena.
A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248 245
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A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
2.2.2. Selection of components for e6aluation Components for seismic safety evaluation were the same as those described in Section 2.1.1 above, namely safety related components, reliability related components and other important components. 2.2.3. Selection of aging degradation for e6aluation The aging degradation phenomena that should be considered in the seismic safety evaluation were selected from the selected phenomena in Section 2.1.4 above through the screening as follows: 1. The aging degradation phenomena that do not affect the seismic related destruction modes, such as the insulation degradation of electric components. 2. The aging degradation that cannot occur as long as the existing maintenance program will be conducted. For example, the thinning of structures by the general corrosion cannot occur, provided that the structures are coated with paint and the painting will be appropriately maintained. 3. The aging degradation that is judged technically not to affect seismic safety. For example, the thinning of the internal surface of valves by general corrosion does not affect seismic safety because it has sufficient wall thickness with the adequate margin for the long-term plant operation. 2.2.4. Seismic safety e6aluation procedure The seismic safety evaluation was carried out according to the following procedure. 1. Select components to be evaluated corresponding to the aging degradation phenomena which are chosen in accordance with Section 2.2.3. 2. Calculate the seismic force in accordance with the seismic classification. 3. Combine the seismic force with other loads, such as inner pressure, etc. 4. Conduct the seismic response analysis with the loads, which are prescribed in ‘Technical Guideline for a Seismic Design of Nuclear
Power Plants’ (JEAG-4601, 1984), considering the aging degradation after 60-year operation, such as the wall thinning by erosion, irradiation embrittlement, cracking by IGSCC, etc. 5. Compare the calculated stress with the allowable stress of the material.
3. Results of Part 2 study The Part 2 study basically confirmed that there is no problem in continuing the existing maintenance program against the aging degradation for the 60-year operation. Throughout the Part 2 study, some items were recommended to improve the maintenance methods for the long-term operation. They should be reflected in the existing maintenance program (see Table 1). Some examples of these improvements are as follows: For the RPV internals, the present material is susceptible to IGSCC, so the scheduled inspections using a TV monitor is required. And the preventive measures, including replacement of core shroud, should be considered. For the primary containment vessel, the planned measurement against the plate thickness at special parts of the PCV is required. For the turbine inner-casing, the material of the casing is carbon steel which is subject to erosion, so the measurement of the plate thickness at typical parts of the casing is required. For the anchor bolts, which are common to every component, they cannot be inspected as they are buried in the concrete. So, it is necessary to conduct sampling inspections on proper occasions. The conservative seismic evaluation was conducted for the aging degraded components and it was confirmed that the calculated stress was less than the allowable stress of the material. Therefore, it was concluded that, for earthquake resistance, there was no problem.
Erosion
Corrosion
Inner-casing
Anchor bolts (common)
Turbine
Corrosion
The water chamber and shell
Heat exchangers (regenerative H/T in CUW)
IGSCC
Shroud support
PVC steel plates Corrosion
IGSCC
Core shroud
RPV internals
Primary containment vessel
Aging
Component
No problem with the parts of the bolts buried in concrete
The material is carbon steel, which is subject to erosion
The material is carbon steel, which is subject to corrosion
PCV was designed with no corrosion margin, so we have to manage the wall thickness
The material is high-nickel alloy that has lower IGSCC sensitivity than stainless-steel
The present materials are susceptible to IGSCC
Long-term integrity evaluation
Table 1 Part 2 evaluation results for Fukushima-1 Unit No. 1 (Example)
Visual inspection and penetrant inspection (at welding part) The bolts cannot be inspected
Leak test, but no other inspection has been carried out since replacement
Whole and local leak rate tests and painting film management
Visual inspections were conducted by an under water TV camera
Visual inspections were conducted by an underwater TV camera
Existing maintenance service
Conducting sampling inspections on proper occasions
Measuring the plate thickness at typical parts of the casing
In future, its integrity will have to be checked
Measuring the plate thickness at typical parts of the PCV
Continued scheduled inspections and planning to take preventive measures including replacement Continuing scheduled inspections and improving the environment by HWC
Measures against aging degradation
A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248 247
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A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
4. Conclusion In Japan, adequate inspections and maintenance are applied to the three oldest LWR plants which have been in operation for over 29 years. In addition, adequate preventive measures are applied to these plants, based on the detailed surveys of the operating experiences from other plants both in domestic and foreign countries and the latest available technical knowledge. In this study, important considerations for conducting improved maintenance activities as part of the plant life management for the three oldest Japanese LWR plants were identified on the assumption that the plants would be retained in service for 60 years.
.
From now on, the evaluation results of the Part 2 study are going to be reflected in the existing maintenance program for the three LWR plants. The new maintenance program will be implemented from the first annual inspection, after each of these plants reaches 30 years of commercial operation. Reevaluation will also be made at 10-year intervals thereafter. We intend to apply the plant life management activities to the next generation LWR plants before they reach their 30th year of commercial operation. It is also important to develop the improved inspection and maintenance methods based on the results of this study.
Technical note
Plant life management activities of LWR plants in Japan Akiyoshi Minematsu * Department of Nuclear Power Engineering, Engineering R&D Di6ision, Tokyo Electric Power Company, 1 -1 -3 Uchisaiwai-cho, Chiyoda-ku, Tokyo 100 -0011, Japan Received 22 February 2000; accepted 30 August 2000
Abstract Plant life management activities of Japanese LWR plants have been conducted since the early 1990s by the utilities and MITI (Ministry of International Trade and Industry) cooperatively. In Japan, where the regulatory practices are different from those in the US, there is neither law nor regulation that prescribes a licensed plant life for nuclear power plants. When an annual inspection is completed without any problem, the next cycle of operation would be permitted and this cycle can be repeated. However, it is generally known that mechanical components and structures deteriorate as they get older. So, we consider it very important to evaluate the long-term integrity of major systems, structures and components of old nuclear power plants. Japanese plant life management study consists of two parts. Both parts of the study were carried out confirming the integrity for the long-term operation of the three oldest Japanese LWR plants: Tsuruga Power Station Unit No.1 (BWR), Mihama Power Station Unit No.1 (PWR) and Fukushima Dai-ichi Nuclear Power Station Unit No.1 (BWR). The Part 1 study was conducted for the purpose of obtaining an outlook for long-term safety operation and was completed in 1996. The Part 2 study was conducted ensuring the plant integrity for the long-term operation in terms of, not only safety, but also reliability. The results of the Part 2 study were made public in February, 1999. Then, the recommended maintenance items were to be added to the existing maintenance programs of the three LWR plants. © 2001 Published by Elsevier Science B.V. Keywords: Plant life management; Power plants; Power stations
1. Plant life management activities
1.1. Introduction
Presented at the 26th Water Reactor Safety Meeting, Bethesda, Maryland, USA, October 26-28 1998. * Tel.: +81-3-42164901; fax: + 81-3-35968544. E-mail address: [email protected] (A. Minematsu).
Already more than 29 years have passed since the first Japanese nuclear power plant went into operation. In Japan, nuclear power plants, both new and old have been operated and maintained in a similar way.
0029-5493/01/$ - see front matter © 2001 Published by Elsevier Science B.V. PII: S 0 0 2 9 - 5 4 9 3 ( 0 0 ) 0 0 3 8 6 - 1
242
A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
In the early 1990s, as Europe and the US were establishing their technical studies on the plant life management and the plant life extension of nuclear power plants, Japan has also come to recognize the importance of plant life management activities. This paper briefly describes the current status of the activities of the plant life management activities of LWR plants in Japan and outlines the future plans.
1.2. Basic concept The Part 1 study was designed to work out the concept of the plant life management through evaluation to confirm the integrity of the major components during the long-term operation of nuclear power plants. In the actual evaluation, 60-year operation was assumed. For the Part 1 study, the three oldest LWR plants described above were selected as the representative plants. Seven or nine components that were the most important for safety plant operation and difficult to repair or replace were selected for the Part 1 technical evaluation. Those were, reactor pressure vessel, reactor internals, primary containment vessel etc. For the Part 2 study, we reviewed the wider range of components (amounting to thousands of items) not only from the safety point of view, but also from the perspective of avoiding an unplanned shutdown for the three oldest LWR plants. This part of the study was intended to develop measures against aging degradation and establish the evaluation technique for the plant integrity during the long-term operation.
60-year operation properly. For example, in the case of BWR plants, it was recognized that the planned measurement against the wall thickness of thinned carbon steel was necessary. Furthermore, it is necessary to have the scheduled inspections of the reactor internals whose material is not low carbon stainless steel to avoid damage due to IGSCC. In the case of PWR plants, it was recognized to be necessary that measures would be taken against stress corrosion cracking of the parts made of alloy 600, such as the steam generator tubes. However, it was confirmed that the plant integrity could be preserved basically by the continuity of the present maintenance methods.
2. Part 2 evaluation method
2.1. Procedure of the Part 2 e6aluation The evaluation methods used in the Part 2 study are as follows (the outline of the evaluation methods is as illustrated in Fig. 1).
2.1.1. Selection of components for e6aluation The purpose of the Part 2 study was to evaluate the long-term plant integrity in terms of safety and reliability. In accordance with this purpose, the components to be evaluated were selected and were classified into three categories as follows. 2.1.1.1. Safety related components. This category includes the components falling under PS-1, PS-2, MS-1 and MS-2 categories in the guideline2:
1.3. Results of the Part 1 study In the Part 1 study, the integrity evaluations were conducted for seven components (BWR case), those were reactor pressure vessel, reactor internals, primary containment vessel, primary coolant piping, primary loop recirculation pump, cable and concrete structure. Through the Part 1 technical evaluation, additional inspections or monitoring were recommended for some of the components to ensure the
2 ‘The Guidelines for Classification of Important Components for Light Water Reactor Facilities’, published by the Japanese Nuclear Safety Commission (NSC), provides the following definitions: PS, abnormality prevention system: loss of the function of the PS systems may subject the reactor facilities into the abnormal condition and thereby cause excessive radiation exposure to the general public or personnel; MS, abnormality mitigation (or suppression) system. The spread of abnormality in the reactor facilities will be prevented or put under control by the MS systems promptly, thereby preventing or suppressing the excessive radiation exposure to the general public or personnel.
A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
In the case of BWR — reactor pressure vessel, high pressure core injection pump, main steam isolation valve, feedwater piping, emergency diesel generator, air conditioning system of the central control room, etc. In the case of PWR — reactor pressure vessel, heat removal pump, pressure relief valve of pressurizer, auxiliary feedwater piping, emergency diesel generator, sea water pump, etc.
2.1.1.2. Reliability related components. This category includes components related to avoiding an unplanned shutdown, that fall under the PS-3 category in the guideline1: In the case of BWR — turbine, generator, generator excitation system, condensate system (including main condenser), feedwater system, circulating water system, main transformer, switching yard, etc. In the case of PWR — turbine, generator, generator excitation system, condensate system (including main condenser), extraction steam piping, main transformer, switching yard, etc.
243
2.1.1.3. Other important components. This category means all components which belong to the guideline except the safety and reliability related components. In the case of other important components, there are not as many differences between BWR and PWR. This category includes, fire protection system, radwaste disposal system, house boiler system, sea water intake system, etc.
2.1.2. Classification and selection of representati6e components In the case of BWR, the components to be evaluated amount to : 3000 items, according to the above selection criteria. In order to conduct the evaluation rationally, the grouping method was introduced. First, the selected components, as noted previously, were sorted into the component categories like pumps, motors etc.
Fig. 1. PLM evaluation procedure in Part 2 study.
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A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
Secondly, the components were further divided into groups according to grouping criteria, such as type, structure, operation circumstances (which include operated location, fluid properties, etc.) and materials. Thirdly, the representative components were selected from each group based on their importance, operating conditions (pressure, temperature) and other criteria.
2.1.3. E6aluation The representative components selected at Section 2.1.2 were broken down to the parts level and technically evaluated by considering the aging degradation phenomena. The evaluations for the representative components were conducted based on the experimental data, the actual plant operational experience etc. Combining these evaluation results and the current inspection and maintenance program of the components, the synthetic evaluation was carried out for the long-term operation. In case the current inspection and maintenance program of the component are judged to be insufficient, additional inspection and maintenance items should be recommended. The evaluation results for the representative components were applied to the other components which belong to the same group, considering the differences between the components. 2.1.4. Selection process of the aging degradation phenomena to be considered There are a lot of aging degradation phenomena in nuclear power plants. It is impractical to consider every possible phenomenon for each part. Therefore, the screening process was adopted to select those phenomena in terms of the environmental condition, the materials etc. The aging degradation phenomena that should be considered in this evaluation were selected as follows (see Fig. 2). 2.1.4.1. Compiling phenomena which occur in industrial materials and products. The aging degradation phenomena that were commonly found to occur in industrial materials and products (both electrical and mechanical) were selected. These phenomena were identified based on the survey of
literature, the study of past abnormalities and the latest technological information.
2.1.4.2. First phase screening of phenomena with respect to nuclear power plant en6ironment. The aging degradation phenomena that do not occur under the condition of light water reactors could be excluded. For example, it is not necessary to consider the halogenated corrosion (necessary to consider in the case of chemical plants) and the sigma phase embrittlement (necessary to consider in the case of temperature between 565 and 930°C). 2.1.4.3. Second phase screening of phenomena with respect to material en6ironment. The aging degradation phenomena that do not occur under the material environment could be excluded. For example, the carbon steel corrosion does not need to be considered when it is used in the anti-corrosive water quality environment. Also, it is not necessary to consider the thermal aging embrittlement for casting stainless steel used at low temperature. 2.1.4.4. Third phase screening of phenomena with respect to the experience. The aging degradation phenomena that can be justified technically to occur infrequently or with very little possibilities under the operating condition based on material test data, could be excluded. In the screening process, operational records and special factors that are specific to the plants should be taken into account. For example, it is not necessary to consider the inter-granular stress corrosion cracking of the austenitic stainless steel (304 stainless steel) under the environment of low concentration of dissolved oxygen (below 8 ppm) and low temperature (B 100°C). 2.2. Seismic safety e6aluation considering aging degradation 2.2.1. Purpose of e6aluation It is important to ensure whether the aged components remain sound or not under seismic loads.
Fig. 2. Principles of identifying possible aging degradation phenomena.
A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248 245
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A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
2.2.2. Selection of components for e6aluation Components for seismic safety evaluation were the same as those described in Section 2.1.1 above, namely safety related components, reliability related components and other important components. 2.2.3. Selection of aging degradation for e6aluation The aging degradation phenomena that should be considered in the seismic safety evaluation were selected from the selected phenomena in Section 2.1.4 above through the screening as follows: 1. The aging degradation phenomena that do not affect the seismic related destruction modes, such as the insulation degradation of electric components. 2. The aging degradation that cannot occur as long as the existing maintenance program will be conducted. For example, the thinning of structures by the general corrosion cannot occur, provided that the structures are coated with paint and the painting will be appropriately maintained. 3. The aging degradation that is judged technically not to affect seismic safety. For example, the thinning of the internal surface of valves by general corrosion does not affect seismic safety because it has sufficient wall thickness with the adequate margin for the long-term plant operation. 2.2.4. Seismic safety e6aluation procedure The seismic safety evaluation was carried out according to the following procedure. 1. Select components to be evaluated corresponding to the aging degradation phenomena which are chosen in accordance with Section 2.2.3. 2. Calculate the seismic force in accordance with the seismic classification. 3. Combine the seismic force with other loads, such as inner pressure, etc. 4. Conduct the seismic response analysis with the loads, which are prescribed in ‘Technical Guideline for a Seismic Design of Nuclear
Power Plants’ (JEAG-4601, 1984), considering the aging degradation after 60-year operation, such as the wall thinning by erosion, irradiation embrittlement, cracking by IGSCC, etc. 5. Compare the calculated stress with the allowable stress of the material.
3. Results of Part 2 study The Part 2 study basically confirmed that there is no problem in continuing the existing maintenance program against the aging degradation for the 60-year operation. Throughout the Part 2 study, some items were recommended to improve the maintenance methods for the long-term operation. They should be reflected in the existing maintenance program (see Table 1). Some examples of these improvements are as follows: For the RPV internals, the present material is susceptible to IGSCC, so the scheduled inspections using a TV monitor is required. And the preventive measures, including replacement of core shroud, should be considered. For the primary containment vessel, the planned measurement against the plate thickness at special parts of the PCV is required. For the turbine inner-casing, the material of the casing is carbon steel which is subject to erosion, so the measurement of the plate thickness at typical parts of the casing is required. For the anchor bolts, which are common to every component, they cannot be inspected as they are buried in the concrete. So, it is necessary to conduct sampling inspections on proper occasions. The conservative seismic evaluation was conducted for the aging degraded components and it was confirmed that the calculated stress was less than the allowable stress of the material. Therefore, it was concluded that, for earthquake resistance, there was no problem.
Erosion
Corrosion
Inner-casing
Anchor bolts (common)
Turbine
Corrosion
The water chamber and shell
Heat exchangers (regenerative H/T in CUW)
IGSCC
Shroud support
PVC steel plates Corrosion
IGSCC
Core shroud
RPV internals
Primary containment vessel
Aging
Component
No problem with the parts of the bolts buried in concrete
The material is carbon steel, which is subject to erosion
The material is carbon steel, which is subject to corrosion
PCV was designed with no corrosion margin, so we have to manage the wall thickness
The material is high-nickel alloy that has lower IGSCC sensitivity than stainless-steel
The present materials are susceptible to IGSCC
Long-term integrity evaluation
Table 1 Part 2 evaluation results for Fukushima-1 Unit No. 1 (Example)
Visual inspection and penetrant inspection (at welding part) The bolts cannot be inspected
Leak test, but no other inspection has been carried out since replacement
Whole and local leak rate tests and painting film management
Visual inspections were conducted by an under water TV camera
Visual inspections were conducted by an underwater TV camera
Existing maintenance service
Conducting sampling inspections on proper occasions
Measuring the plate thickness at typical parts of the casing
In future, its integrity will have to be checked
Measuring the plate thickness at typical parts of the PCV
Continued scheduled inspections and planning to take preventive measures including replacement Continuing scheduled inspections and improving the environment by HWC
Measures against aging degradation
A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248 247
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A. Minematsu / Nuclear Engineering and Design 207 (2001) 241–248
4. Conclusion In Japan, adequate inspections and maintenance are applied to the three oldest LWR plants which have been in operation for over 29 years. In addition, adequate preventive measures are applied to these plants, based on the detailed surveys of the operating experiences from other plants both in domestic and foreign countries and the latest available technical knowledge. In this study, important considerations for conducting improved maintenance activities as part of the plant life management for the three oldest Japanese LWR plants were identified on the assumption that the plants would be retained in service for 60 years.
.
From now on, the evaluation results of the Part 2 study are going to be reflected in the existing maintenance program for the three LWR plants. The new maintenance program will be implemented from the first annual inspection, after each of these plants reaches 30 years of commercial operation. Reevaluation will also be made at 10-year intervals thereafter. We intend to apply the plant life management activities to the next generation LWR plants before they reach their 30th year of commercial operation. It is also important to develop the improved inspection and maintenance methods based on the results of this study.