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Western Pressurized Water Reactors (PWRs) are considering the replacement from lithium hydroxide (LiOH) to potassium hydroxide (KOH) as a pH balancing agent in primary water due to the higher cost and lower availability of 7Li. This study qualified the irradiation-assisted stress corrosion cracking (IASCC) propagation compatibility of baffle plate materials in simulated PWR primary water with LiOH or KOH chemistries. Samples extracted from two pieces of Type 304 stainless steel baffle plates with different dose (9 and 15 dpa) were investigated. Water chemistries were switched between LiOH (1.3–2 ppm) and KOH (7.2 to 11.3 ppm) at constant pH during the crack growth rate (CGR) test on a sample, thus the effect of water chemistry was determined directly without other variations. Results indicate that there is no discernable effect on the CGR behavior by changing from LiOH to KOH.

Investigations of Type 347 and Type 316 austenitic stainless steel baffle-former bolt failures at 4-loop downflow pressurized water reactors (PWRs) in the US have identified significant amounts of transgranular cracking on the fracture surfaces along with intergranular cracking and ductile rupture. The expected failure mode for baffle-former bolts is irradiation-assisted stress corrosion cracking (IASCC), which typically results in an intergranular cracking mode. The presence of transgranular cracking raised the question of whether IASCC was responsible for the failures, or if a fatigue component was contributory. Detailed fractography and metallography of bolts with and without cracks confirmed that an environmentally-assisted cracking mechanism was responsible. Reviews of the literature identified numerous cases where stress corrosion cracks (including irradiation-assisted stress corrosion cracks) in austenitic stainless steels transitioned from intergranular to transgranular, typically as a more severe stress state developed. The potential cracking mechanisms are discussed. This allowed the conclusion that IASCC was responsible for the degradation observed in the baffle-former bolts, though a component of alternating loading cannot be excluded. The downflow plant configuration results in increased stress on neighboring intact bolts after bolt failure, resulting in the development of ‘clusters’ of failed bolts, which grew as additional bolts failed. This cluster growth contributed to the severe stress state imposed on the bolts.

The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.