Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-17
Abstract： [Background]: The volume of the nuclear power plant containment is huge, making it difficult to conduct equal-scale or large-scale thermal-hydraulic tests. Currently the test data mainly come from small-scale tests. [Purpose]:To address the applicability of small-scale containment test data in validation process of the containment performance analysis code, the analysis method for applicability of experimental data is proposed and developed on the basis of similarity analysis of the pressure response process in the containment. [Methods]: The applicability study of the test data, which are produced by some scaled containment facilities such as the HDR, Battelle and CVTR, is carried out in combination with the test parameters. The applicability of each test case are obtained respectively when they are applied to validate the containment code in case of the Large Break Loss of Coolant Accident (LBLOCA) and Main Steam Line Break Accident (MSLB) of HPR1000 nuclear power plant. [Results]:The results show that the similarity criteria for pressure response process and key phenomena within the containment vessel under accident conditions can be used to analyze the applicability of different containment tests to the target power plant. [Conclusions]:The proper combination of test cases including HDR ISP-16&23, Battelle CASP-1&2, and CVTR T3 can represent the pressure transient process, results of coupling phenomena such as mass and energy release at the break, condensation near the containment shell and internals, within the HPR1000 containment in case of LOCA or MSLB. The distortion is either within the acceptable range or conservative for design limits of containment pressure, so that the small-scale containment test data are suitable for the verification and validation of the HPR1000 containment thermal hydraulic response analysis code.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-03
Abstract： [Background]: The volume of the nuclear power plant containment is huge, making it difficult to conduct equal-scale or large-scale thermal-hydraulic tests. Currently the test data mainly come from small-scale tests. [Purpose]:To address the applicability of small-scale containment test data in validation process of the containment performance analysis code, the analysis method for applicability of experimental data is proposed and developed on the basis of similarity analysis of the pressure response process in the containment. [Methods]: The applicability study of the test data, which are produced by some scaled containment facilities such as the HDR, Battelle and CVTR, is carried out in combination with the test parameters. The applicability of each test case are obtained respectively when they are applied to validate the containment code in case of the Large Break Loss of Coolant Accident (LBLOCA) and Main Steam Line Break Accident (MSLB) of HPR1000 nuclear power plant. [Results]:The results show that the similarity criteria for pressure response process and key phenomena within the containment vessel under accident conditions can be used to analyze the applicability of different containment tests to the target power plant. [Conclusions]:The proper combination of test cases including HDR ISP-16&23, Battelle CASP-1&2, and CVTR T3 can represent the pressure transient process, results of coupling phenomena such as mass and energy release at the break, condensation near the containment shell and internals, within the HPR1000 containment in case of LOCA or MSLB. The distortion is either within the acceptable range or conservative for design limits of containment pressure, so that the small-scale containment test data are suitable for the verification and validation of the HPR1000 containment thermal hydraulic response analysis code.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-23
Abstract： In-situ leaching is the main process of natural uranium production in China. The environmental protection problems in the process of construction, operation and decommissioning have been widely concerned. The production of wastes was introduced in this paper, and ten issues of concern were raised, included bleed ratio, drilling mud disposal, monitoring well setting, groundwater baseline determination, abnormal data determination of monitoring well, wastewater treatment, evaporation pool operation, target value determination of groundwater restoration, groundwater remediation and waste minimizaton. the present situation of technology and management was discussed, and the main task and research direction in the field of environmental protection of in-situ leaching of uranium mining were pointed out, in order to provide some suggestions for the sustainable development of in-situ leaching of uranium mining。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-07
Abstract：蒸汽发生器传热管是压水堆核电站一回路的关键压力边界，正常运行及事故条件下的系统运行参数，直接影响着传热管结构的完整性。为进一步对传热管的结构响应特性进行研究，建立了两种主流排列方式的传热管局部模型，并分别进行了正常运行和典型事故参数下的流固耦合计算和对比分析，获得了一、二次侧流体共同作用下的传热管关键部位的载荷变化规律。结果表明，在相同的流动条件下，传热管叉排模型的应力和变形均大于顺排模型；传热管根部横截面的等效应力沿壁厚方向先减小再增大，中部横截面的等效应力沿壁厚方向呈现近似线性降低；在传热管根部区域，内、外壁面的等效应力沿轴向高度先迅速降低而后快速升高，并形成应力平台。
Subjects: Nuclear Science and Technology >> Engineering of Nuclear Power submitted time 2023-07-04
Abstract： It provides a necessary reference for the study of MPWR serious accident integration analysis program development standard and the development of serious accident integration analysis procedure. Based on the current domestic regulatory requirements and regulations, software development regulations, software development, nuclear safety analysis and design software evaluation requirements, studied in this paper, suggested the standard system, quality assurance, verification and confirmation, and preliminarily established the standard system of serious accident program development. Based on the practice of existing procedure evaluation, the evaluation requirements of serious accident procedure and its focus are discussed. On the basis of the existing and ongoing soft development and component evaluation of safety analysis of nuclear power plant, the standard system of the development of serious accident procedures has been preliminarily established, and preliminary suggestions have been made for the development specification of integrated analysis procedures for serious accidents of pressurized water reactor.
Peer Review Status:
Awaiting Review
Subjects: Other Disciplines >> Synthetic discipline submitted time 2023-03-28 Cooperative journals: 《中国科学院院刊》
Abstract： China has formed a comprehensive nuclear industry system, but the radioactive waste disposal capacity cannot keep up with the requirement since more wastes are generated and accumulated. One of most critical challenges is the safety, environmental, and even political issues caused by the overdue storage of radioactive waste. In this study, the radioactive waste management experience in foreign countries, which have high nuclear power installed capacity and rich experience in radioactive waste disposal, has been investigated and analyzed, And then the comparison has been made between these countries and China, in terms of the policy and legislative system, regulatory and implementing, financing and compensation. Finally, the implications on China’s governance in the radioactive waste disposal have also been presented in the paper.
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