Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against non-ductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities and institutes voluntarily participating as members. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches and results.
JSME rules for fitness for service have flaw acceptance rules for cast austenitic stainless steel (CASS) pipes. They allow applying two-parameter and elastic-plastic fracture mechanics methods using Z-factor. However they do not clearly describe whether limit load method is applicable for the case of no or low thermal aging condition. The authors performed tensile fracture tests using flat plate specimens with a surface flaw and confirmed that limit load method is applicable in the conditions of no thermal aging and even fully saturated thermal aging with high ferrite number. Also the plate with a shallow flaw ruptured at the critical stress defined by nominal stress at rupture-flaw depth curve in the code case which was determined by the similar flat plate tests of stainless steel or nickel alloy specimens. These results will be reflected to the revision of the code.
The UOi+ fluorescence spectra were measured for solutions whose conditions, except for coexisting species, corresponded to high level waste (HLW) solutions in the nuclear fuel reprocessing plants. From these spectra it was ascertained that the U concentration can be determined in the range of 2-200 g/m3 with an accuracy of 1 g/m3 at 30--40°C. The UO;' fluorescence was influenced significantly by the temperature. The apparant activation energy of the temperature effect was 35.7 kJ/mol. The fluorescence intensities increased with nitric acid concentration in the range of 0.75-4.5 mol/dm3. The temperature effect was considered to be caused by a chemical process concerned with water molecules.
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