The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI welding residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material. This paper discusses Phase III of the WRS characterization program, comparing measured and predicted weld residual stresses profiles through the dissimilar metal weld region of pressurizer safety and relief nozzles removed from a cancelled plant in the United States. The DM weld had already been completed on all of the plant nozzles before use in the mock-up program. One of the nozzles was completed with the application of the stainless steel safe-end weld to a section of stainless steel pipe. Measurements were taken on the nozzles with and without the welded pipe section. Several independent finite element analysis predictions were made of the stress state in the DM weld. This paper compares the predicted stresses to those found by through-thickness measurement techniques (Deep Hole Drilling and Contour Method). Comparisons of analysis results with experimental data will allow the NRC staff to develop unbiased measures of uncertainties in weld residual stress predictions with the goal of developing assurances that the analysis predictions are defensible through the blind validation provided using well controlled mock-ups and ex-plant material in this program.

This paper provides a technical basis for a crack growth rate (CGR) for use in performing evaluations of cracking in stainless steel canister materials. The source of crack initiation and growth is deposition of chloride aerosols on the canister surface followed by deliquescence leading to a brine solution. The brine solution attacks the stainless steel surface, leading to pitting; in the presence of tensile stress (such as residual tensile stress due to welding), stress corrosion cracking can occur. The CGR will be used for evaluating flaw growth under a proposed ASME Boiler and Pressure Vessel Code Case covering stainless steel canisters.

The residual stresses imparted by the welding process are a principal factor in the process of primary water stress corrosion cracking (PWSCC) of Alloy 82/182 nickel-alloy (i.e., dissimilar metal or DM) piping butt welds in PWRs. While Section XI of the ASME Code requires that residual stresses are considered in crack growth calculations, there is little guidance or requirement on how to calculate them. Analytical models are frequently used to simulate the welding process in order to predict the residual stress distribution in the weld and base material as an input to crack growth calculations. The crack growth calculations, in turn, have demonstrated a high sensitivity to the welding residual stress distribution inputs. While significant progress has been made in understanding and reducing the variability in calculated residual stress among modelers as well as the variability in measured residual stress among different techniques, there remains some uncertainty regarding any given measured or calculated distribution. A feasible alternative to calculating through-wall stress distributions with analytical models on a case-by-case basis is to develop a set of standardized through-wall stress distributions that are applicable to DM welds. Examples of standardized through-wall distributions for residual stress are found in numerous consensus code and standards. The benefit of established through-wall stress distributions is that evaluations for flaws in welds would start from a uniform basis on one of the key inputs to the crack growth calculation, reducing the time required to perform and review flaw evaluations. This paper presents and describes the technical basis for a set of through-wall distributions for common DM welds found in the US nuclear industry. The basis of the distributions include the results of analytical models, including uncertainty, as well as measured data for through-wall stress in DM welds.

A probabilistic model was developed that considers the likelihood of through-wall penetration of chloride-induced stress corrosion cracking (CISCC) in austenitic stainless steel canisters and compares different population-based sample inspection regimes. This paper describes the inputs and methods used to simulate multiple canisters with a range of susceptibilities. This paper also summarizes results of key illustrative cases.

ASME Boiler and Pressure Vessel Code Section XI Code Case N-860 provides inspection requirements and evaluation standard for welded stainless steel canisters used for spent nuclear fuel storage. The Code Case defines an initial inspection interval and populations then defines examination requirements that are based on the primary degradation mechanism, chloride induced stress corrosion cracking (CISCC). Additional examination requirements are based on the results of the initial screening examination. This paper summarizes the technical basis for the examinations and the evaluation criteria defined in Code Case N-860. Technical basis information for topics related to the inservice inspection and the flaw evaluation of canisters are described. The topics related to inservice inspection include: 1) the reasons for and the basis of requirements for the site susceptibility of the canister installations, 2) the inspection intervals required by the Code Case, 3) the inspection sample population required by the Code Case, 4) the methods and acceptance criteria for visual examinations required by the Code Case, and 5) the size and location of the required inspection region for supplemental examinations. The topics related to the flaw evaluation include: 1) a summary of the crack growth rate technical basis, and 2) background related to flaw size evaluation for spent fuel canisters.