福岛第一核电厂事故源项估算及方法比较

本文参考日本福岛第一核电厂的部分资料,利用美国核管会发布的《轻水堆核电厂事故源项》（NUREG-1465）以及国际原子能机构发布的《为轻水堆设计估算参考源项所提供的简化方法》（IAEA-TECDOC-1127）两份技术文件中的假设条件,分别计算出事故后由堆芯释放到安全壳内的放射性源项。同时通过对堆芯积存量、抑压水池净化...     

乏燃料事故源项估算方法研究

福岛事故之后乏燃料的安全问题受到广泛关注。本文介绍一自主开发的应急辅助决策软件STEM中的乏燃料事故源项估算模块。利用美国核管理委员会(NRC)开发的核电事故后果分析软件RASCAL 4.2与STEM分别对假定事故情景进行计算,结果验证了STEM的正确性。STEM乏燃料事故源项估算模块可为核电厂的乏燃料事故后果评价提供参考。     

RASCAL及其在核事故后果评价中的应用

介绍了美国核管会用于核与辐射事故后果分析的辐射评价系统(RASCAL)的主要功能和特性,重点分析了RASCAL的源项计算剂量模块、场外监测数据计算剂量模块、气象数据处理模块,以及源项计算模式、大气输运扩散模式和剂量计算模式。最后,将RASCAL应用于我国某核电厂事故应急演习中,评价分析事故情景下的放射性影响,并将其结果通过Google Earth进行三维展示。     

核电厂大破口失水事故(LBLOCA)始发严重事故释放源项的分析

本文利用一体化的严重事故数据计算分析程序,研究核电厂发生大破口失水(LBLOCA)事故始发严重事故情况下裂变产物的释放、迁移、去除和最终在不同区域的分布等特征。假设核电厂具有双层安全壳设计并且安全壳保持完整性的情况下,计算最终向环境的释放源项。最后利用美国核管会(NRC)的NUREG-1465假设的壳内事故源项的释放份额计算环境释放源项的份额,并对结果进行比较。计算结果可以为应急设施评价源项的选取以及场外后果评价提供参考。     

核动力厂选址假想事故源项的探讨

介绍了美国核审管部门对核动力厂选址假想事故源项确定的历史演变以及相关研究进展。推荐了一套核动力厂选址假想事故源项,以及确定非居住区边界的评价分析方法和假设条件。计算了M310和AP1000核动力厂厂址的非居住区的最大径向距离。计算结果表明:(1)对于M310和AP1000核动力厂,采用确定论方法估算大气扩散因子,安全壳泄漏率(体积分数)取0.3%/d,非居住区最大径向距离分别不超过1.4 km和1.9 km;(2)在厂址选址阶段,考虑核动力厂的技术路线尚未明确或采用新技术路线,采用"不考虑安全壳喷淋系统等能动安全设施对核素的去除作用"的选址假想事故源项来确定厂址非居住区的边界基本上是可行的。     

美国核管会削减研究预算和项目

【英国《国际核工程》1984年10月号第3页报道】由于研究预算削减和源项研究方面数以百万美元计的变动,美国核管理委员会计划在1985财政年度终止在爱达荷国家工     

ARCON96程序理论模型分析

大气弥散因子是评价核电厂控制室可居留性的重要参数,美国核管理委员会采用ARCON96程序评估该参数。对ARCON96程序的基本原理及主要理论模型进行分析,并以安全壳表面释放情况为例,对程序中的面源释放扩散模型开展敏感性分析。为ARCON96程序的科学使用提供建议,保障计算的合理性。     

泳池式常压低温供热堆燃料转运跌落事故源项计算分析

正《泳池式常压低温供热堆运行及事故工况分类》中的所有事故分析结论表明,乏燃料转运跌落事故源项可以包络其他事故源项。事故源项计算基本假设为:1)反应堆运行时间为5a,冷却时间为10d,乏燃料在从堆水池到乏燃料水池的过程中发生事故,使用ORIGEN2     

压水堆释放源项快速估算程序开发

事故时向环境释放的源项是确定核电厂(NPP)应急响应水平和防护行动决策的重要依据。基于电厂工况估算源项是核电厂严重事故应急响应期间重要的应急评价内容之一。在国际原子能机构(IAEA)和美国核管会(NRC)的有关技术文档基础上,本文介绍了基于压水反应堆(PWR)工况进行事故释放源项估算的步骤和基础数据,并归纳了7种实用的事故释放源项估算方法。基于这些方法,开发了PWR事故时环境释放源项快速估算程序。该程序为不同估算方法提供4种释放途径:安全壳泄漏、安全壳旁通、蒸汽发生器传热管破裂(SGTR)和直接环境释放,除直接环境释放途径外,其他释放途径都估算了核素释放过程中的衰变、滞留、喷淋和过滤等减弱过程。对比发现,软件计算结果与美国核管会的RASCAL软件释放源项计算结果接近。     

美国核管理委员会在源项问题上持谨慎态度

【英国《国际核工程》1985年10月号第3页报道】美国核管理委员会已建议在法规管理领域中谨慎地应用源项研究的新成果。 1985年8月核管会散发了供讨论的核管会源项再评价计划进展报告(Nureg 0956),报告介绍了用来进行法规管理实践再评价的新的分析方法,这些方法的基础是反应堆安     

Allowable stresses to limit shearing deformation in reinforced concrete containment design

An extensive program of the U.S. Nuclear Regulatory Commission (NRC) to study reinforced concrete containment wall behavior has been completed for orthogonal reinforcement. The transfer of shear caused by the action of seismic load has been studied sufficiently to recommend the seismic shear design and allowable shear stresses. However, the recommendations made in this paper are not the NRC position for the design.     

Temporary non-code repair of ASME Code Class 3 piping

When a flaw is detected in piping, it may be impractical to perform a repair in accordance with Section XI of the ASME Code because the plant may have to be shut down. The U.S. regulations require licensees of nuclear power plants to obtain written relief from the U.S. Nuclear Regulatory Commission (NRC) before a repair other than that specified in Section XI is performed. On June 15, 1990, the NRC issued Generic Letter (GL) 90-05 providing evaluation guidelines and acceptance limits for proposed non-code temporary repairs of ASME Code Class 3 piping when a code repair is impractical. Flaw evaluation, type of repairs, monitoring, augmented inspection, and the term that temporary repairs are applicable are all addressed in GL 90-05. The guidelines in GL 90-05, as modified by experience learned from its implementation, are under consideration as a draft ASME Code Case.     

美国联邦法规10 CFR 50附录Ⅰ的修订方案及对我国辐射防护审管的启示

本文详细介绍了美国核管理委员会(NRC)对轻水堆的设计目标基准的修订方案和策略,并在此基础上,考虑到我国核电厂址向内陆发展所致公众照射途径的变化,提出了需要明确核动力厂设计目标值的建议,以及应用现行辐射防护相关标准需要关注的问题:(1)ICRP第103号出版物从以前基于过程的实践和干预的方法发展为基于辐射照射情况性质(计划照射、应急照射和现存照射)的方法,应当注意区分不同的照射情况;(2)ICRP第103号出版物在数值上更新了当量剂量和有效剂量的辐射权重因子和组织权重因子,因此,实施剂量评估所采用的剂量转换因子也需要更新。     

Performance-goal based (risk informed) approach for establishing the SSE site specific response spectrum for future nuclear power plants

This paper provides the technical basis for the performance-goal based approach presented in the American Society of Civil Engineering Standard ASCE/SEI 43-05 for establishing the safe shutdown earthquake (SSE) site specific design response spectrum (SSRS) for future nuclear power plants. This approach has been adopted by the U.S. Nuclear Regulatory Commission (NRC) in their Regulatory Guide 1.208. This approach has now been followed at more than 20 sites.In addition to more thoroughly documenting the basis for the ASCE 43-05 performance-goal based approach, the paper also summarizes the minimum seismic core damage frequency (SCDF) achieved at 28 Central and Eastern U.S. sites when the SSRS is defined by this approach coupled with other NRC requirements. The minimum SCDF lies in the range of less than 6 × 10⁻⁶/yr to 0.6 × 10⁻⁶/yr with a median ratio of about 3 × 10⁻⁶/yr.     

核电厂信息安全问题研究及建议

信息和网络技术在核电厂的广泛应用,给核安全带来了新的威胁与挑战。本文深入分析了核电厂的信息安全现状以及面临的主要威胁,总结归纳了IAEA和美国NRC针对核电系统信息安全的研究成果、具体措施以及法规标准,最后结合中国的实际情况分析提出了关于保障我国核电厂信息安全的若干建议,为我国进行核电厂信息安全相关的决策及政策法规研究提供借鉴。     

Loose-parts monitoring: Present status of the technology, its implementation in U.S. reactors, and some recommendations for achieving improved performance

At the request of the U.S. Nuclear Regulatory Commission (NRC), an assessment of the technical development status of loose-parts monitoring systems (LPMS) and their performance record to date in commercial light-water-cooled nuclear reactor plants was made during the spring of 1977, using an on-site personal interview and equipment demonstration approach. Our study revealed that while presently demonstrated LPMS technology does indeed provide a capability for detecting the presence of those relatively massive loose parts that would likely constitute a serious operational or safety hazard to the plant, it unfortunately affords little information useful to the determination of the parts' safety significance and has not yet attained the levels of sophistication and reliability ordinarily associated with safety systems. We also found a definite need for specification of the functional requirements for LPMS, in the form of a clear and comprehensive statement of NRC policy regarding the formulation and implementation of safety-oriented, yet operationally practicable, loose-parts monitoring programs for both existing and future nuclear generating stations so that overall objectives of both the utilities and the regulatory agency might be satisfied simultaneously.

While it is our best technical judgment that loose-parts monitoring programs providing reliable detection (but not characterization) capabilities could be implemented with today's technology, the path on which the nuclear utility industry should proceed in order to meet NRC expectations is not completely clear. A Regulatory Guide entitled “Loose Part Detection Program for the Primary System of Light-Water-Cooled Reactors,” soon to be issued for public comment, constitutes a first step towards satisfying this need for guidance and goal establishment.     

Evaluation of nuclear facility decommissioning projects program-status

Major studies have been undertaken in recent years by the US Nuclear Regulatory Commission (NRC) and others on the technology, safety and costs associated with decommissioning nuclear facilities. The program described in this presentation is being undertaken by the NRC to compile and evaluate the activities of ongoing decommissioning projects. Assessment and evaluation of the methods, impacts, and costs will provide a basis for evaluating licensee's decommissioning proposals and for future decommissioning direction and regulation.Program participants include the US Nuclear Regulatory Commission (NRC) through the Office of Regulatory Research, UNC Nuclear Industries (UNC) through the Decommissioning Programs Department, and nuclear facility licensees.     

Integration of nondestructive examination reliability and fracture mechanics

A multi-year program on the Integration of Nondestructive Examination and Fracture Mechanics (NDE/FM) has been funded by the U.S. Nuclear Regulatory Commission at the Pacific Northwest Laboratory. Many activities are being pursued under this program. This paper highlights some of the activities: input to the NRC Pipe Crack Task Group, an evaluation of manual ultrasonic testing of centrifugally cast stainless steel, interaction matrix, advanced UT technique evaluation, qualification document, evaluation of crack characterization techniques, international NDE reliability work, siamese imaging technique for imaging planar-type radial defects in reactor piping, fracture mechanics analysis for PTS-type flaws and piping reliability, and a position paper on piping ISI.     

Piping dynamic reliability and code rule change recommendations

The conservative nuclear piping design criteria for seismic and dynamic loads have led to piping systems with excessive numbers of snubbers. To improve this undesirable situation, a Piping and Fitting Dynamic Reliability Program was initiated by the Electric Power Research Institute (EPRI) in 1985 with cooperation from the U.S. Nuclear Regulatory Commission (NRC). The objective of the program is to develop improved, realistic, and defensible ASME design rules by taking advantage of the inherent dynamic margins in the nuclear piping system. The research results have demonstrated that piping systems have large reserve dynamic capacity and the dynamic failure mode is due to fatigue or fatigue-ratcheting rather than plastic collapse. Based on such physical evidence, a set of code rule change recommendations is suggested in its preliminary form.