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In order to ensure safety and efficient state of nuclear power plant in the long time operation, it is necessary to figure out the acceptance table of defects of the primary loop to meet the requirement of rapid assessment in engineering. It is necessary to do the fracture mechanics and plastic limit load analysis for the cracked pipe. Through numerical simulation analysis, the rationality of adopting plastic limit load as the critical state of austenitic stainless steel structure with good fracture toughness is demonstrated. Based on the criterion of non-decreasing the plastic limit load in design state, the acceptance standard of defects for the austenitic stainless steel pipe is established in this paper. The results show that the defect acceptance size of the primary circuit pipe is close to that in the ASME code, but much larger than that of the RSE-M code, and the RSE-M code is too conservative.

In the engineering of aerospace and nuclear power engineering, the safety and reliability of small pipe（Outer diameter less than 2-inch, about 50 mm） has become one of the main factors restricting the safe operation. However, in the existing design codes, there is a lack of detailed technical requirements and evaluation criteria for small pipeline. In this paper, based on the experimental and FEA analysis, the fatigue and final fracture of pipeline components are discussed. The results show that the results of the fatigue crack propagation by experimental and FEA method are same, and they account about 3 % of the fatigue life in the study case. The critical failure mode is the plastic collapse, so it is to increase the critical crack size by increasing the tensile strength.

Firstly, the experimental J-R curve and the predicting curves by the three models of 400℃ thermally aged Z3 CN20-09 M pipe and elbow were compared. The results showed that ANL model and H3 T model aren’t suitable for the prediction of low toughness CASS components, however the EDF model is better for low toughness CASS components prediction. All the three models present non-conservative prediction for the early crack propagation, but it doesn’t affect the threshold judgment. According to the comparison between experimental J-R curves and the predicting curves, the ANL model and H3 T model are more suitable for the thermal aging assessment of CASS Z3 CN20-09 M.

The pressurized thermal shock（ PTS） event poses a potentially significant challenge to the structural integrity of the reactor pressure vessel（ RPV） during the long time operation. In most countries（ with exception of the USA）,proving RPV integrity under the PTS load is based on the deterministic fracture mechanics（ DFM）. In the USA,the "screening criteria"for maximum allowable embrittlement of RPV material,which form part of the USA regulations,are based on the probabilistic fracture mechanics（ PFM）. In this paper,the FAVOR software of Oka Ridge National Laboratory is used to compare the difference between PFM and DFM analysis of the PTS even. The analysis method of FAVOR is shown through case study of the benchmark transient in the IAEA-TECDOC-1627 report. Finally,the limit of the current FAVOR（ Version6. 1） is shown.