The primary objective of this report is to evaluate materials degradation issue unique to the operational environments of LWSMR. Concerns for specific primary system components and materials are identified based on the review of design information shared by mPower and NuScale. Direct comparisons are made to materials issues recognized for advanced large PWRs and research activities are recommended as needed. The issues identified are intended to improve the capability of industry to evaluate the significance of any degradation that might occur during long-term LWSMR operation and by extension affect the importance of future supporting R&D.The evaluations documented in this report highlight that neither large advanced PWR designs, nor LWSMR primary system designs, introduce significant changes in material selection or fabrication processes from those being applied in Gen. II reactor plant component repairs and replacements. Although differences in configuration introduced by LWSMRs create some unique circumstances that warrant study, in most areas the new materials R&D needed to support the application is not fundamentally distinct from current LWR research programs, but rather represents a possible extension of these programs. To address LWSMRs, R&D plans should be reviewed and adjusted as needed to ensure the relevant design configurations, environmental conditions and performance expectations are addressed.An example of a new design configuration that suggests a need for materials R&D is the use of a helical coil steam generator with primary coolant flowing on the tube OD and secondary coolant boiling occurring on the tube ID surfaces. In this case, the broad areas of R&D suggested relate to ensuring no SCC vulnerabilities are introduced by fabrication processes needed for the helical coil tube bundle and the need for improvements in thermo-hydraulic modeling capabilities. A less obvious example of a new design configuration relates to vessel fabrication practices. Vessel fabrication will certainly be more complicated due to the integration of all of the primary system components from a traditional PWR into a single vessel enclosure.Examples of new environmental exposure conditions include reactor vessel fluence and CRD operating environment. The smaller diameter and lower operating pressures used by LWSMR designs allow for significantly thinner vessel shells, but with higher EOL neutron fluence. As a consequence, significant radiation damage occurs through a greater fraction of the wall thickness. With regard to CRDs and CRD penetrations, some LWSMRs will locate the CRDs at the top of the integrated vessel, causing them to be exposed to steam at higher pressurizer temperatures.As significant changes in material selection are unlikely for LWSMR designs, research to resolve key materials degradation concerns identified for large advanced PWRs remains of high importance and expanded activities are needed in many areas. Significant benefit for LWSMRs can be gained by R&D to characterize the effects of component fa...

As the existing light water reactor (LWR) fleet ages, the weldability of structural materials used to construct the reactor pressure vessels (RPVs) and reactor internals is diminished. The decrease in the weldability in austenitic and ferritic materials is attributed to the formation of helium in the material microstructure. Helium (He) generation occurs during the service life of irradiated reactor internals from neutron transmutation reactions of boron and nickel in these materials. Welding on irradiated materials, if performed without appropriate consideration of fluence exposure and helium generation, can result in a heat affected zone cracking phenomenon termed helium induced cracking (HeIC). The heat input associated with welding is a major factor affecting the coalescence of the generated helium along grain boundaries. As the material cools, the tensile stresses generated from welding can cause cracking to occur along grain boundaries weakened by helium bubble coalescence. In some cases, the preferred or only method of repair or replacement of a reactor internal component is welding. For components located in regions of low thermal fluence, the welding process implementation may be relatively straightforward and only heat input control may be required. However, in high thermal fluence regions, weld repair of irradiated reactor internal components is complicated by the presence of high concentrations of helium and significant care must be taken in welding process selection and heat input control. This paper highlights envisioned applications for weld repair on irradiated reactor internals. It also summarizes recently completed guidance published by the EPRI Materials Reliability Program (MRP) and EPRI BWR Vessel and Internals Project (BWRVIP) which provides an improved basis for plants to assess the weldability of components at various locations within the reactor.

Recent inspections have identified cracking in the core shroud that is atypical in that the cracks exhibit characteristics inconsistent with traditionally reported intergranular stress corrosion cracking (IGSCC) occurring within stainless steel weld heat-affected zones (HAZs). These flaws are oriented transverse to the weld and are observed to propagate significantly beyond the weld HAZ. This paper describes the investigations which have been performed to quantify the likely limits on growth of these “off-axis” SCC cracks. The investigation includes welding residual stress analysis to determine the stress field present adjacent to the weld and crack tip SIF calculations for the bivariant stress field.