各有关单位等离子体物理研究所.docx

1、各有关单位等离子体物理研究所一、 反应堆和核电厂H-084-001水堆燃料芯块-包壳相互作用Pellet-clad interaction in water reactor fuels.R,e/INIS-FR-08-0305, 2004, 26p.001H-084-002先进热工水力与中子编码现状和未来应用专题讨论会会议录Proceedings of the workshop on advanced thermal-hydraulic and neutronic code: current and future applications.R,e/ NEA-CSNI-R-2001-2, 2001

2、, 723p.001H-084-003用核动力工程设计试验中心沸水堆全尺寸细网格束试验模拟来评价COBRA-IE计算机编码Cobra-IE evaluation by simulation of the NUPEC BWR full-size fine-mesh bundle test (BFBT).R,e/B-T-3653, 2006, 16p. 001H-084-004在瑞典核电站会见中反映出的安全管理特征：系统观点方法Safety management characteristic reflected in interviews at Swedish nuclear power plan

3、ts: a system perspective approach. R,e/SKI-R-07- 25, 2005, 46p.001H-084-005概率安全目标：第1阶段情况和瑞典与芬兰的经验Probabilistic safety. Phase 1 status and experiences in Sweden and Finland. R,e/SKI-R-07-06, 2007, 66p.001H-084-006国际研究堆大会：安全管理及有效利用（详细文摘）International conference on research reactors: safe management an

4、d effective utilization. Book of extended synopses. R,e/IAEA-CN-156, 2007, 297p.001H-084-007应急柴油发电机共因故障收集与分析的国际共因故障数据交换项目报告ICDE project report on collection and analysis of common-cause failures of emergency diesel generators.R,e/ NEA-CSNI-R-2000-20, 2001, 78p.001H-084-008容器内堆芯退化编码验证矩阵更新（1996- 1999年

5、）（OECD/NEA专家组报告）In-vessel core degradation code validation matrix update 1996-1999. report by an OECD/NEA group of experts. R,e/NEA-CSNI-R-2000-21, 2001, 351p. 001H-084-009第43号国际标准问题：编码验证用快速硼稀释瞬态试验（比较报告）International standard problem (ISP) No.43. rapid boron-dilution transient tests for code verific

6、ation. Comparison report.R,e/ NEA-CSNI-R-2000-22, 2001, 197p.001H-084-010第41号国际标准问题：根据放射性碘试验设施实验练习安全壳碘计算机编码International standard problem (ISP) No.41. containment iodine computer code exercise based on a radioiodine test facility (RTF) experiment.R,e/ NEA- CSNI-R-2000-6-VOL-1, 2000, 174p.001H-084-01

7、1在长期前景中短期严重事故管理行动的影响最终报告Impact of short-term severe accident management actions in a long-term perspective. Final report.R,e/ NEA- CSNI-R-2000-8, 2000, 16p.001H-084-012凭严重事故管理深入了解安全壳中碘、铯、锶和其它裂变产物的释放控制Insights into the control of the release of iodine, cesium, strontium and other fission products in

8、the containment by severe accident management.R,e/ NEA- CSNI-R-2000-9, 2000, 110p.001H-084-013国际共因故障数据交换项目报告：电动阀共因故障收集与分析ICDE project report: collection and analysis of common-cause failures of motor operated valves.R,e/ NEA-CSNI- R-2001-10, 2001, 44p.001H-084-014使用核电厂安全性能指标的摘要报告Summary report on th

9、e use of plant safety performance indicators.R,e/ NEA- CSNI-R-2001-11, 2001, 26p.001H-084-015以法国原子能委员会管道弯曲试验为基础的疲劳裂纹增长基准技术报告Technical report on the fatigue crack growth benchmark based on CEA pipe bending tests.R,e/ NEA-CSNI-R- 2001- 14, 2001, 53p.001H-084-016评估水水动力堆失水事故与瞬态热工-水力编码的验证矩阵：OECD水水动力堆热工-水

10、力编码验证矩阵支持组报告Validation matrix for the assessment of thermal-hydraulic codes for VVER LOCA and transients. A report by the OECD support group on the VVER thermal- hydraulic code validation matrix. R,e /NEA-CSNI-R-2001-4, 2001, 249p.001H-084-017退化堆芯问题情况：G. Bandini与NEA退化堆芯冷却任务组合作编写的综合论文Status of degrad

11、ed core issues. Synthesis paper prepared by G. bandini in collaboration with the NEA test group on degraded core cooling.R,e/ NEA- CSNI-R-2001-5, 2001, 15p.001H-084-018先进热工-水力与中子编码的现状和未来应用（摘要及结论）Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions.

12、 R,e/ NEA-CSNI-R-2001-9, 2001, 79p. 001H-084-019依赖关系分析指南：共因故障分析北欧/德国工作组第1阶段项目报告：试验情况比较与应用Dependency analysis guidance. Nordic/ German working group on common cause failure analysis. Phase 1 project report: comparisons and application to test cases.R,e/SKI-R-07-41, 2007, 42p. 001H-084-020北欧核电站“R-书”中管

13、系部件的可靠性数据（项目第1阶段）（第1版）Reliability data for piping components in Nordic nuclear power plant R-book. project phase 1. Rev.1.R,e/SKI-R- 08-01, 2008, 100p.001H-084-021液态金属冷却堆：设计和运行经验Liquid metal cooled reactors: experience in design and operation.R,e /IAEA-TECDOC-1569, 2007, 272p.001H-084-022第11届原子能研究专题

14、讨论会分文会议录Proceedings of the eleventh symposium of atomic energy research.R,e/INIS-SK- 2008-003, 2001, 853p.001H-084-023第17届原子能研究专题讨论会会议录，第1卷Proceedings of the seventeenth symposium of atomic energy research. Vol. I.R,e/INIS-SK-2008-001, 2007, 592p.001H-084-024第17届原子能研究专题讨论会会议录，第2卷Proceedings of the s

15、eventeenth symposium of atomic energy research. Vol. II.R,e/INIS-SK-2008-108, 2007, 492p. 001H-084-025全部核设施地震再评价专题讨论会会议录Proceedings of the workshop on the seismic re-evaluation of all nuclear facilities.R,e/NEA-CSNI-R-2001-13, 2001, 435p.001H-084-026实施严重事故管理措施专题讨论会会议录Proceedings of the workshop on t

16、he implementation of severe accident management measures.R,e/NEA-CSNI- R-2001-20, 2001, 456p.001H-084-027亚洲核合作论坛2006年研究堆利用（合同研究）专题讨论会Proceedings of the FNCA 2006 workshop on the utilization of research reactor (contract research).R,e/JAEA-Conf- 2007-008, 2007, 326p.001H-084-028JRR-3堆堆芯燃耗计算方法Core bur

17、n-up calculation method of JRR-3. R,e/JAEA-Conf-2007-008, p.167-208, 001H-084-029JRR-3堆中用MVP和MVP-BURN编码进行计算Calculation using MVP and MVP-BURN in JRR-3.R,e/ JAEA-Conf-2007-008, p. 209-221,001H-084-030高通量先进中子应用堆中子学计算系统综述Overview of the neutronics calculation system for the HANARO.R,e/ JAEA- Conf-2007-

18、008, p.222-225,001H-084-031“快堆循环工艺开发项目”2006财年研发活动评估报告（期中报告）Assessment report of research and development activities in FY2006 activity.Fast reactor cycle technology development project (Interium report)R,j/JAEA- Evalua- tion-2007-003, 2007, 100p.001H-084-032核电厂老化的社会-经济、卫生及环境影响Ageing of power plants

19、socio-economical, sanitary and environmental impact.R,f /INIS-FR-08-0787, 2005, 139p.001H-084-033国际建立新的人可靠性评估专题讨论会会议录：系统误差的研究到应用Proceedings of the international workshop on building the new HRA: errors of commission-from research to application.R,e/NEA-CSNI-R-2002-3, 2003, 270p.001H-084-034加速器驱动次临界研

20、究堆的中子学设计Neutronic design of an accelerator driven sub-critical research reactor. R,e/INIS-RS-1384, 2002, 8p.001H-084-035高温工程试验堆无人值班乏燃料流监测核保障系统的发展Development of the unattended spent fuel flow monitoring safeguards system for the high temperature engineering test reactor.R,e/JAEA-Technology-2007- 003,

21、 2007, 31p.001H-084-036高温工程试验堆提升功率试验以来的运行经验Operating experiences since rise-to-power test in high temperature engineering test reactor.R,e/JAEA- Technology-2007-014, 2007, 69p.001H-084-037采用低缩铀硅化物燃料的JRR-4的堆芯特点：初始堆芯及燃耗堆芯Core characteristics of JRR-4 using low-enriched-uranium-silicide fuel. Initial c

22、ore and burn-up core.R,j/ JAEA-Technology-2007-017, 2007, 101p. 001H-084-038JRR-4堆运行实践和反应堆物理实验指南Guidance of operation practice and reactor physics experiments using JRR-4. R,j/ JAEA-Technology-2007-018, 2007, 114p. 001H-084-039高温工程试验堆产氢系统：模拟试验设施的结构和主要规格HTTR hydrogen production system. Structure and

23、main specifications of mock-up test facility. R,j/ JAEA- Technology-2007-022, 2007, 219p.001H-084-040通过实际环境的应用来减少第4代核能系统建设费用（最终报告）Generation IV nuclear energy systems construction cost reductions through the use of virtual environments- final Report.R,e/DOE-SF-22327-Final, 2005, 67p.001H-084-041核反应堆

24、与脱盐过程结合的最佳化Contribution to the optimization of the nuclear reactors to desalination processes.R,f/FRNC-TH-7327, 2007, 131p.001H-084-042高温工程试验堆气体压缩机油密封性能的改进Improvement in oil seal performance of gas compressor in HTTR.R,j/JAEA- Technology-2007-047, 2007 48p.001H-084-043JRR-3堆硅化物燃料堆芯的反应性管理和燃耗管理Reactiv

25、ity management and burn-up management on JRR-3 silicide-fuel core. R,j/JAEA- Technology-2007-050, 2007 48p.001H-084-044JRR-3堆安全保护系统的回路精确度Loop accuracy of JRR-3 safety protection system. R,j/JAEA-Technology-2007- 052, 2007 56p.001H-084-045安全性能指标专家会议会议录Proceedings of the specialist meeting on safety p

26、erformance indicators.R,e/ NEA-CSNI-R-2002-2, 2002, 345p.001H-084-046第19届中子散射在凝聚物质研究中应用专题讨论会：会议时间表和报告摘要The XIX workshop on neutron scattering application for condensed matter investigations program and summaries of reports.R,ru/INIS-RU-508, 2006, 108p.001H-084-047国际原子能机构的中子数据汇编Neutron data compilati

27、on at the International Atomic Energy Agency. R,e/INDC(NDS)-0001, 1968, 7p.001H-084-048在燃烧等离子体中氢同位素与轻元素的堆芯浓度（IAEA顾问会议摘要报告）Core concentrations of hydrogen isotope. and light elements in burning plasma Summary report of an IAEA consultants meeting.R,e/INDC(NDS)-0518, 2007, 14p.001H-084-049IAEA国际核反应数据中

28、心网络技术会议报告Report on the IAEA technical meeting of the international network of nuclear reactor data centers.R,e/INDC(NDS)- 0519, 2007, 96p.001H-084-050聚变堆重元素杂质的原子数据：第2届研究协作会议摘要报告Atomic data for heavy element impurities in fusion reactors. Summary report of the second research coordination meeting.R,e

29、/INDC(NDS)-0521, 2008, 25p.001H-084-051计算非能核应用核反应参数第3届研究协作会议摘要报告（参考输入参数数据库：第3阶段）Summary report of third research coordination meeting on parameters for calculation of nuclear reactions relevance to non-energy nuclear application (Reference input parameter library: phase III).R,e/INDC(NDS)- 0524, 200

30、8, 46p.001H-084-052FENDL-2.1：聚变用评价核数据库的更新FENDL-2.1: update of an evaluated nuclear data library for fusion applications.R,e/ INDC(NDS)-0467, 2004, 34p.001H-084-053次锕系元素核反应数据咨询会议摘要报告Summary report of consultants meeting on minor actinide nuclear reaction data. R,e/INDC(NDS)-0512, 2007, 21p. 001H-084-

31、054等离子模拟用原子与分数数据：第2届IAEA研究协作会议摘要报告Atomic and molecular data for plasma modelling. Summary report of second IAEA research coordination meeting.R,e/ INDC(NDS)-0515, 2007, 22p.001H-084-055聚变堆中氚存量：最终研究协作会议摘要报告Tritium inventory in fusion reactors. Summary report of the final research coordination meeting.R,e/ INDC(NDS) -0516, 2007, 30p.001H-084-056优先原子与分子数据评估：IAEA技术会议摘要报告Assessme