In this section we look at applied management practice around the world.

74.1 Degradation of reactor vessels

There have been no cases of ageing degradation of reactor vessels. Thermal annealing at high temperature (475 ± 15°C) for 100-150 h has been applied to plants operating in Russia since 1987 as a preventative measure (Badanin, 1989; Cole and Friderichs, 1991).

74.2 Degradation of reactor internals

There is a report that a guide tube support pin made of Alloy 750 was dam­aged in Mihama Nuclear plant in Japan in 1978. The damaged pin was mov­ing around as a foreign body and it was discovered in the steam generator chamber. It was determined that the damage was from high stress in pri­mary water at high temperature. After this incident, research studies were conducted over many years to understand the damage mechanisms. It was found that Alloy 750 was a material sensitive to PWSCC, and that sensitivity increased largely depending on heat treatments. Since it was replaced with a new cold worked material (CW316 SS), no additional cracks have been found.

Regarding baffle former bolts, some cracks were found in an old French power plant (Fessenheim and Bugey) in 1988. Some bolts had 10-25 dpa fluence after being in operation for 10-20 years. They were made of a 316 cold worked stainless steel and showed intergranular stress corrosion crack­ing (IGSCC). The ageing mechanism was assumed to be IASCC. The cracks had spread from the shank zone of the head to the lower part of the head. The material was found to be hardened and radiation-induced segregation was found in the grain boundary. According to the hardness profile mea­surement, it was approximately 5-10 dpa, and there was no evidence of swelling. The bolts with cracks were detected at rows 2 and 3 from the lower part of the nuclear reactor where a considerable amount of neutron radi­ation had accumulated. According to the report, until that point, cracks in the baffle former bolts had been found in the ‘down-flow’ design in which inlet coolant flows downward (Gerard, 2009). From 1989 to 1993, the flow of the coolant in the nuclear reactor of the CP0 (name of French PWR 900 MW pre-series units) plant was changed to up-flow and, between 2000 and 2003, one third of the bolts were replaced. The cracking rate of baffle bolts increases slowly depending on dose. Based on the information on all the baffle bolts researched during in-service inspection (ISI) for all CPO units, the dose threshold is estimated to be approximately 3-4 dpa. The number of cracked bolts increases slightly at higher doses.

In Japan, two kinds of approach have been applied to the PWR plants since 1998 in light of the baffle former bolt issues experienced in France and also in the United States. The first approach was to replace the baffle former bolts. From 2001 to 2002, type 347 stainless steel bolts were replaced with cold worked 316CW stainless steel bolts in Mihama Unit 1 and 2. The second approach was to replace the internal structure of the nuclear reac­tor. In this case, the lower zone including the baffle former bolts and the upper zone were replaced. Since 2004, the internals of the nuclear reactors in three PWR plants have been replaced entirely. The measures described above were applied to the baffle former bolts in a 2-loop PWR plant which was built in the early 1970s.

A research programme aimed at preventing defects of the internals of nuclear reactors has been ongoing since 2000. For example, there has been research on IASCC of austenitic stainless steel used in PWR inter­nals such as in baffle former bolts. Valuable data was collected through this research, which showed that IASCC initiation is closely related to stress and exposed neutron fluence. In other words, the threshold stress which deter­mines IASCC occurrence is dependent on the amount of neutron fluence. The threshold stress value tends to decrease as neutron fluence increases.

Based on the data and experiments, it was decided that a guideline would be published with detailed interpretation. Japan published a guideline on management activities such as inspection of the baffle former bolts of actual power plants in 2002. This guideline will be revised in the future to reflect up-to-date knowledge obtained through international collaboration.

In Korea in 2007, defects were found in the control rod guide tubes made of Inconel X-750 in the internals of a nuclear reactor. The control rod guide tubes were replaced with CW 316 stainless steel tubes. Since then, no cracks have been found in the baffle former bolts (Hwang et al, 2010).

In the United States, the Westinghouse Owners Group (WOG) has inspected cracks in baffle former bolts from PWRs in other countries. They have also provided information on activities planned for Westinghouse power plants with potential cracking. The WOG has clearly demonstrated that destruction of minor bolts would not have a serious impact on safety because a number of baffle former bolts would still support the structure. The WOG activities are as follows:

• Development of analytical methods and acceptance criteria for bolt analysis.

• Performance of risk-informed evaluations.

The Nuclear Energy Institute (NEI) which consists of the Materials Technical Advisory Group (MTAG) of the United States is formed of energy company representatives. The MTAG has received support from EPRI to prepare a guideline for In-Service Inspection (ISI) of RVI equipment which has signif­icant impact on continued and safe operation of power plants. In preparing an inspection guideline, the damage to equipment inside the nuclear reactor by inspection, fatigue, abrasion and corrosion were considered. Recently, Westinghouse and AREVA have published a report using screening based on various significant damage mechanisms as a part of EPRI MRP on reac­tor internals.