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[en] ASME Code procedure for evaluating the acceptability of flaws detected during in-service inspection is revised. Critical crack size for instability is proposed as criteria for detected flaws in operating plants

[en] Program materials were three weldments fabricated from A533 Grade B class 1 plate material and Mn Mo Ni weld wire. Specimens fabricated from the three submerged arc weldments included Type A Charpy V-notch impact, small size tensile, and 1/2T compact tension specimens. After encapsulation, the specimens were irradiated at the UVAR to two fluence levels, 8 x 10¹⁸ n/cm² and 1.5 x 10¹⁹ n/cm² (E > 1 MeV). Specimens were subjected to sequences of irradiation and anneals and then tested. Metallurgial/mechanistic analyses were also performed. It was concluded that excellent recovery of all properties could be achieved by annealing at greater than or equal to 850⁰F (454⁰C) for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the expected rate. Several drawbacks were identified for wet thermal annealing. A conceptual dry in-situ thermal annealing procedure was developed for thermal annealing embrittled reactor vessels

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[en] Westinghouse has utilized precipitation hardening nickel base alloys for many years under reactor operating environment in a variety of applications. Inconel 600, 718 and X-750 has been used for coil springs, hold down springs, and in general all types of springs in the reactor internals. In addition, these alloys have been used as core support clevis material, all types of internal bolting and for guide tube support pins. Westinghouse used these nickel base precipitation hardening alloys for many years without an indication of stress corrosion cracking (SCC) phenomenon. It wasn't until 1978 that the first instance of SCC failure of Inconel X-750 occurred in a Westinghouse designed nuclear steam supply system. The failures of Inconel X-750 guide tube support pins prompted extensive material characterization and stress corrosion cracking test programs throughout the world. Westinghouse's testing program resulted in major changes in the Materials Specification for procurement of Inconel X-750. The changes included chemistry control, a well defined heat treatment, and closer control on grain size and micro-structure. This paper presents the evolution of material specifications as they pertain to inconel X-750. Test results that suggest that testing in steam at 400 degree C can be utilized to rank precipitation hardening alloys as to their susceptibility to SCC, and recommendations as to other alloy systems for reactor internals fasteners and bolting. 18 figs

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[en] An EPRI sponsored program was carried out by Westinghouse to determine the extent of fracture toughness recovery as a function of annealing time and temperature for neutron embrittlement sensitive reactor vessel material and to develop an optimal thermal anneal procedure for field applications. Program materials were three weldments fabricated by Combustion Engineering, Inc., from the same heat of A533 Grade B Class 1 plate material and the same heat of MnMoNi weld wire. The only variables were the target copper level and the welding flux which was Linde Grade 80 and Linde 0091. Weldments of 0.22, 0.36, and 0.41 wt % copper were produced. It was concluded from this study that excellent recovery of all properties could be achieved by annealing at 850₀F (454₀C) and above for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the rate which would have been expected had no anneal been performed. System limitations were identified for both wet and dry annealing methods

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[en] The feasibility and methodology for thermal annealing an embrittled reactor vessel was investigated with the primary goal being to develop an in-situ thermal annealing procedure which would maximize fracture toughness recovery, minimize re-exposure sensitivity, and minimize reactor downtime. The feasibility was assessed by evaluating whether plant design and system limitations (reactor vessel design, metallurgical, reactor vessel insulation, primary shield concrete, RCS piping and associated equipment support, and reactor vessel internals and fuel storage) would be exceeded by an imposed localized 850⁰F heat source. In addition, health physics problems were also considered. Finally, a conceptual thermal annealing apparatus was developed and a general thermal annealing procedure was established. The results of this work indicate that an anneal temperature of 850⁰F for a holding period of 168 hours, obtained by using resistance-type heating elements attached to a support apparatus, could be used in an effort to restore the fracture toughness properties of ferritic materials of the reactor vessel as required by 10CFR Part 50 Appendix G