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Chapter 5 Operation and Maintenance of PWR Plants

  1. Maintenance to Cope with Plant Aging

As of the end of 2007, 23 PWR plants were operating in Japan, 9 of which will have been operating more than 30 years by 2010.

Components and systems of NPPs have often been found degraded by mechanisms associated with plant aging, such as thinning, SCC, and degrading of electrical insulating material. These degradations of components and systems are serious concerns as they worsen and expand their affected areas and eventually degrade functions of plant facilities.

To cope with plant aging effects and ensure plant safety, it is necessary to take long-sighted views to properly evaluate potential aging effects on plant systems and components and to establish preventive maintenance plans.

  1. Technical Evaluation of Aging Effects

Countermeasures to plant aging effects have been implemented based on the “Basic Approaches to Cope with Plant Aging” published by the Agency for Natural Resources and Energy, In April 1996. In October 2003, the following requirements were published by the agency as regulatory requirements.

  1. Within 30 years after the start of plant operation, the utility should technically evaluate aging effects on the plant and establish a long-term (10-year) plant maintenance plan based on the evaluation results.

ii )The utility should reevaluate the results of the technical evaluations and the long-term maintenance plan at periodical safety reviews of the plant performed every 10 years, and re­establish a long-term maintenance plan for the following 10 years.

Technical evaluations of aging effects on plants and the establishment of long-term maintenance plans based on the evaluations have already been conducted at six domestic PWR units: Mihama Nos.l and 2, Takahama Nos.l and 2, Genkai No.l and Mihama No.3. Review results on the reports of these activities had been published by the agency as of May 2007.

In the technical evaluations on the six plants, possible aging phenomena to which countermeasures should be considered and

their possible effects on plants within a 60-year assumed plant life are listed on component and structure bases. Additional maintenance activities which should be implemented to cope with the plant aging are also listed based on the evaluation results. Main items listed in the evaluation reports are summarized as follows: neutron irradiation embrittlement of reactor vessel; inter-granular stress corrosion cracking (IGSCC) of stainless steel; primary water stress corrosion cracking (PWSCC) of nickel based alloys; irradiation assisted stress corrosion cracking (IASCC); low cycle fatigue; high cycle fatigue; erosion and corrosion of piping; degradation of cable insulators; and neutralization of concrete.

For an example, primary water SCC of nickel based Alloy 600 (Inconel®600) has been evaluated based on failures seen in domestic and overseas plants, and the following additional countermeasures are planned to be implemented during the periods for maintenance of these plants.

i inspecting and monitoring: more frequent checks, additional check points (metal welds), improved accuracy of inspections.

  1. Estimating and evaluating: improved accuracy of equations for estimating generation frequencies and progress speeds, clarification of generation and progression mechanisms.

  2. Repairing and replacing: promotion of Alloy 600 replacement with Alloy 690, reduction of residual stresses by peening or other methods.

  1. Maintenance Techniques to Cope with Plant Aging

To effectively and efficiently implement additional countermeasures extracted in the technical evaluations on the plant aging issues as discussed above, new maintenance techniques for inspection, repair and replacement activities must be developed.

Examples of maintenance techniques needed for major plant components are described below.

(1) Maintenance techniques for reactor core internals

Potential aging degradation phenomena of PWR core internals are the IASCC of baffle former bolts and the wear of control rod cluster thimble guide plates. The former reportedly occurred in an

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overseas PWR power plant

In “Guidelines for Check/Evaluation of Reactor Core Internals of PWR Plants” published by the Thermal and Nuclear Power Engineering Society, equations are proposed for the estimation of IASCC failure probabilities of baffle former bolts caused by plant aging. Checks and replacement of the bolts have been planned or implemented in domestic PWR plants based on the guidelines.

  1. Inspection techniques

IASCC failure of a baffle former bolt would occur at the neck of the bolt where stresses are concentrated. However, since the head of the bolt is welded to the baffle plate to prevent the bolt from turning, the bolt neck cannot be visually checked. Hence, the bolt neck is checked by ultrasonic testing. Since the bolts, more than 1000 are used in a 3-loop plant, are in highly radioactive fields due to radiation from surrounding structures having been irradiated by core neutrons, ultrasonic testing of the bolts needs to be conducted quickly using remotely operated testing equipment underwater.

Dedicated inspection equipment compatible with the restrictive environmental conditions has already been developed, verified and applied to actual plants. Also, dedicated ultrasonic testing equipment for core barrel former bolts has been developed, verified and applied.

ii )Techniques for repairing and replacing

Tie replacement of baffle former bolts has been planned and implemented in certain plants, as one of the preventive maintenance activities, reflecting findings from failure experiences overseas.

After welds at bolt heads are cut and removed, and old bolts are loosened and removed, new bolts made of material with better stress levels are screwed in and swaged to prevent their turning or loosening.

Since replacement of new bolts following the removal of old ones could trigger the seizing of bolts caused by defects such as eccentricities, additional tools and equipment needed to be developed for troubleshooting in such occasions. Equipment for the replacement of bolts, which should be operated remotely underwater and quickly, the similar requirements as for the inspection equipment, has been developed,

verified and applied to actual plants.

All the reactor core internals can be replaced as a single assembly. Therefore, countermeasures reflecting the latest findings can be implemented for all aging degradation phenomena assumed to occur in core internals; this enhances their reliabilities. Moreover, the design of core internals can be modified to prepare for an increased number of control rods required for the use of high burnup fuel.

With the removal of all core internals as a single assembly, gaps at interfaces of these components with the reactor vessel need to be precisely controlled. An equipment design for remotely conducting such precise gap control work has already been developed, verified, and applied to actual plant uses.

(2) Maintenance techniques for piping

The two major aging degradation phenomena of piping are high-cycle thermal fatigue and wall thinning. Since there are more than 50 km of main piping in the NPP primary systems, it is important to select piping parts or segments susceptible to degradation by aging based on a full understanding of aging degradation mechanisms of piping, and to verify the soundness of the piping segments by highly accurate inspections, in order to perform efficient preventive maintenance of the piping.

i) Evaluation of high cycle thermal fatigue cracking

The high cycle thermal fatigue of piping is typically seen at locations where hot and cold water flow merge, or where cavitation occurring in closed branches of piping repeatedly generates local thermal stresses in piping walls due to water turbulence in the flows.

Since such high cycle thermal fatigue piping cracking has often been seen in domestic and overseas plants, JSME published the "Guidelines for the Evaluation of High Cycle Thermal Fatigue of Piping” based on research conducted by private firms. The selection and inspection of piping segments susceptible to aging degradation are now conducted by referring to the guidelines and considering feedback on recent problems including such items as cracking at the outlet piping of a heat exchanger used in a domestic plant. Temperatures of piping segments of

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