The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus, for the RPVs of pressurized water reactors (PWRs) expected to experience neutron fluences from 1 to 5×1019 n/cm2 (>1 MeV) after 40y, the exposures will be 2 to 10×1019 n/cm2 after 80y. Additionally, because the recent pressurized thermal shock (PTS) re-evaluation project has resulted in lower average failure probabilities for PWRs , many plants will increase their operating power level which will further increase the neutron flux and the resultant fluence . Even for normal start-up and cool-down transients, the coolant-pressure-temperature (P-T) curve must be below the corresponding stress (from pressure) that could cause fracture for an assumed very large crack size. Fig. 1.2 shows a schematic depiction of a P-T operating envelope progressively decreased by irradiation embrittlement (RTNDT) of a sensitive RPV steel . Since the power reactor surveillance database contains only sparse data higher than 3×1019 n/cm2, the existing embrittlement models, ([for example, the Eason, Odette, Nanstad, Yamamoto (EONY) model in reference 4], are inadequate as predictive tools to those high fluence levels. To obtain data at the high fluences for life extension will require either very long term surveillance data (for which material is now in short supply), or through the use of test reactor experiments which use high neutron fluxes.
Our current understanding of radiation damage mechanisms suggests that it is not appropriate to use highly accelerated test reactor data directly to predict high fluence behavior for RPV operating conditions. Moreover, there is now experimental evidence that phases rich in Ni and Mn do form in irradiated low Cu steels . Because these phases may require a small degree of Cu precipitation to catalyze their nucleation, they may not contribute to hardening and embrittlement until relatively high fluences. The delayed embrittlement caused by these so-called “late-blooming phases” (LBP) may produce an effect that could have serious implications to RPV life extension. As discussed in [2,5], it is important to understand and quantify the composition-flux-fluence-temperature regime in which they evolve, and develop a better quantitative description of their contribution to embrittlement. The potential for late blooming phases emerging in some composition-fluence-temperature-flux regimes could result in severe underestimates of shifts based on current models by up to 50°C or more [2,5]. The mechanisms that cause irradiation-induced embrittlement of RPV steels are discussed in [2,4,5] and will not be discussed further here, except in the context of the mechanisms that take place during the thermal annealing process which are inextricably linked to those which cause the embrittlement.
Various options are possible to mitigate the effects of irradiation embrittlement on the RPV: (1) fuel management schemes can be used to reduce the neutron flux, which reduces the fluence and, therefore, the embrittlement; (2) shielding of critical areas with, e.g., stainless steel can reduce the flux; (3) the emergency core cooling system (ECCS) water can be heated to reduce the thermal shock effects during a PTS event: (4) the RPV could be replaced; (5) various analytical methods, such as the alternative PTS rule in Title 10, U.S. Code of Federal Regulations, Part 50.61a (10CFR50.61a)  can potentially be used to allow for operation with RTPTS values above the screening criteria; (6) mechanically prestressing the beltline region of the RPV by compressive loading with structural bands ; (7) if a weld is the critical area for embrittlement, replace the weld with more resistant material ; (8) thermal annealing to recover fracture toughness of the RPV materials. Fuel management can have only a slight effect on reducing neutron flux, while shielding, although somewhat effective, is expensive. Conditioning of the ECCS water relates only to thermal shock situations, and replacement of the RPV is not considered practicable at this time. The analytical options, although effective, would likely offer relatively short term relief, while the prestressing and weld replacement concepts have not been thoroughly evaluated. Many of these options are discussed by Planman, Pelli and Torronen .
Recovery of the material toughness through thermal annealing is one method of increasing safety margins of the RPV. Thermal annealing involves heating the RPV beltline materials to temperatures ~ 50 to 200°C above the normal operating temperature for about one week, with the amount of recovery increasing with increasing annealing temperature. Two different procedures can be used to perform the thermal anneal, a wet anneal or a dry anneal. A wet anneal is performed with cooling water remaining in the RPV and is limited to the RPV design temperature of 343C. A dry anneal requires removal of the cooling water and internal components and would normally be performed at temperatures in the range of 430-500C. If thermal annealing is considered, then the post-annealing reirradiation response of the steel must also be evaluated. 10CFR50 specifies thermal annealing as a method for recovering the fracture toughness and refers to Regulatory Guide 1.162 (RG 1.162) . RG 1.162 also provides guidance for determining the amount of recovery, the re- embrittlement trend (assumed to occur at the same rate as in the irradiated case) and for establishing post-anneal material properties.
This report provides an introduction to the subject of thermal annealing of reactor pressure vessels and materials, including a summary of experience with actual thermal annealing of power reactors. The report is prepared in satisfaction of Milestone L-11OR040602 Level M3 #M3L11OR04060203 “Complete initial assessment of thermal annealing needs and challenges.”