[en] The overall safety requirements for nuclear element design are summarized on the basis of investigating regulations and standards used in different countries. The specifications for nuclear elements and systems in standard review plan (SRP) published by Nuclear Regulation Committee (NRC) are reviewed. Besides, the design evaluation methods and acceptance criteria are especially analyzed in detail. The safety review processes of Frarnatome Mark-B11 assembly and Westinghouse 16 × 16 NGF assembly are then investigated respectively according to principles and methods specified by NRC. The practices of safety review for fuel assembly engineering application from abroad show that design evaluations for nuclear elements that improved or developed by our nation should be based on existing operating experience and original design safety evaluation results. The special attention should be paid to effects result from design changes. In view of current regulations and standards, the acceptance criteria of safety analysis for fuel assembly developed in our nation could be established by taking the requirements and practices of NRC into account. (authors)

[en] Inadvertent loading error can occur in a reactor core. Peaking factors of the core could be increased significantly due to the misplaced fuel assemblies. This event is a Condition III incident (an infrequent fault). Technically the cases of misplacing assembles is nearly infinite. Some typical cases need to be chosen for the incident analysis. The purpose of analysis is to affirm the ability of the online core monitoring system to detect the inadvertent loading. The method of the analysis is described in this paper. Most of the misloading can be detected by the online core monitoring system and eliminated by rearranging the core pattern. Some cases not detected still satisfy the safety requirements, which the nuclear enthalpy rise factor F_ΔH is less than the design limits. (authors)

[en] The mechanism, precondition and sensitive conditions of the thermal stratification occurred in pressurized surge line was described. Three key factors influencing thermal stratification was comprehensively analyzed, including flow-rate, coolant temperature and pipeline layout. The results show that these key factors can be combined into an entirety to relieve the thermal stratification. (authors)

[en] PIPESTRESS is professional software for mechanical calculation and analysis. Containing standard such as ASME, RCC-M, EN, it can be used in nuclear grade or non-nuclear grade design and check analysis with modeling, calculating and post-analysis module. As used widely in Westinghouse, AREVA and EDF companies, PIPESTRESS can be applied to three-dimensional calculation of structure, supported facility for heat or fatigue analysis. With application of PIPESTRESS software, the nuclear piping system in the nuclear power plant was used as an analysis model to calculate the weight, thermal expansion, seismic and so on. The selection of ASME design standard was used for static evaluation, through the calculation, the stress and the maximum stress ratios under the condition of different stress combinations were gotten. The results were shown that, some parts of the system could meet the ASME standard limit requirements. Stress analysis and evaluation by the application of PIPESTRESS software, which can not only ensure the safety of the pipeline system, and also can optimize the design. The software was played an important role in the construction and operation of a nuclear power plant. (authors)

[en] Based on the problem of operation redundancy, process complexity and low efficiency in the finite element analysis of nuclear power station, caused by using single software to complete the whole process of finite element analysis, the AutoCAD, HyperMesh and ABAQUS software are applied to complete the pre-processing, solving and post-processing of the whole finite element analysis process based on their respective features. The analysis showed that the integrated use of three types of software can raise the clarity of the calculation, take brief and smooth operation, easy to control and amend the analysis date. This method greatly improve the analysis efficiency. (authors)

[en] Since the construction line defining the initial crack extension will cause significant effect on fracture toughness measurement for ductile metals, the background of the construction line was analyzed based on mechanical model of crack-tip stress field. The popular used British standard E1820-08a and Chinese standard GB21143-2007 were compared for their different definition of the construction line in test method. Finally the reasonability of these two kinds of construction lines was discussed coupling with test result application. (authors)

[en] The corresponding three dimensional entity finite element model and the equivalent solid plate finite element model are established aiming at the calculation difficulty brought by the multi-pore construction characteristic amid the supporting plate of reactor core in the stress analysis. This paper aiming the equivalent model in the checking calculation of the Supporting Plate, used ASME specification method to simulate and analysis the applicable condition. The equivalent model adaptability is established by calculating and comparing the maximum membrane bending stress in the area of the key way and the nearby physical parts of the two models. The result shows: the equivalent solid plate model based on ASME specification can be applied in the supporting plate equivalent model under reactor core and the result is conservative, in addition, which can provide a reference for the subsequent stress analysis and audit calculation. (authors)

[en] Aiming at the structural deformation and failure of the fuel assembly under the earthquake, the fuel assembly is analyzed by the simplified method, and the collision load and stress are calculated under the seismic condition. And then the calculated values are compared with the crush load of the framework and the stress limit of the guide tube, thereby evaluating the structural integrity of the fuel assembly, in order to provide a reference for the analysis and calculation of the seismic performance of the fuel assembly structure in the future. (authors)

[en] Background: Leak-before-break technique is widely used in operating and being built nuclear plants, for which net-section-criteria is adopted by ASME to assess flawed pipes under full yielding condition. Purpose: However, net-section-criteria has been proved to overestimate the loading capacity of structure. Methods: For nuclear pipeline with internal surface crack, finite element method is used to simulate its deformation process under internal pressure, and initial plastic failure load is determined based on J integral curves of crack front. Results: Then, calculated value of initial plastic load is compared with theoretical one from ASME code, and the results show that the theoretical value overestimates the capacity load of structure. Conclusions: Finally, applicability of allowable membrane stress for level-A service restriction set by ASME-BPVC-XI code is evaluated. (authors)

[en] Standard fracture toughness measurement was explored on contact tension specimen made of ductile metal S355 used for nuclear plant structure. During the test, plastic collapse instead of crack extension occurred ahead of crack tip with increasing tension load and only local area around crack tip yielded when the whole specimen failed. After that, Gurson model was used to simulate the failure process of contact tension specimen made of S355. The calculation results show that Gurson model can reproduce crack tip collapse well. At last, a universal failure assessment curve was built for S355 by combining finite element simulation with test measurement, and relationship between plastic failure load and crack length was determined. (authors)