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Review of Studies Regarding Risk-Informing Regulations for Normal Operating Transients

机译:关于正常运营瞬变风险信息规定研究的综述

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The current regulations, as set forth by the United States Nuclear Regulatory Commission (USNRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative. During the past decade, the NRC conducted the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project that established a technical basis to support a risk-informed revision to current PTS regulations (10CFR Part 50.61). Once the results of the PTS re-evaluation are incorporated into a revision of the 10 CFR 50.61 guidance on PTS, the technical basis for the fracture toughness that is required to withstand a PTS event (an accidental loading) will differ from the technical basis for the fracture toughness that is required by Appendix G to 10 CFR Part 50 for normal operating transients. The new PTS guidelines will be based on more realistic risk-informed models and inputs whereas the existing Appendix G requirements contain many known conservatisms that place unnecessary burdens on plant operation. Consequently, a goal of this project is to develop technical information supporting a risk-informed revision to the current requirements of Appendix G to 10 CFR 50 in a manner that is consistent with that used to develop a risk-informed revision to the PTS regulations. This research has consisted of the application of the FAVOR computer code for cool-down transients associated with reactor shutdown and the development and application of the FAVOR computer code for heat-up transients associated with reactor startup. This paper provides a brief overview of the current results of this research project.
机译:如美国核监管委员会(USNRC)所提出的,确保轻水核反应堆压力容器(RPV)在经过计划启动(加热)和关闭时保持其结构完整性(酷 - 向下)瞬态在附录G至10 CFR部分50中指定,其通过引用附录G与ASME代码的第Xi部分结合。这些法规的技术基础包含了技术界广泛认可的许多方面,因为技术界不必要地保守。在过去十年中,NRC进行了跨学科加压热冲击(PTS)重新评估项目,该项目建立了一个技术基础,以支持对当前的PTS法规(10CFR部分50.61)的风险明智的修订。一旦PTS重新评估的结果纳入了对PTS的10 CFR 50.61指导的修订,就可以承受PTS事件(意外装载)所需的断裂韧性的技术基础将与技术为基础不同附录G至10 CFR部分50用于正常操作瞬变所需的断裂韧性。新的PTS指南将基于更现实的风险的风险的模型和输入,而现有的附录G要求包含许多已知的保守主义,这些保守主义在植物操作上放置不必要的负担。因此,该项目的目标是以与用于为PTS规则制定风险明智的修订的方式,开发支持风险明智的修订的技术信息。该研究组成的是,有利于电脑代码的用于与反应堆关闭相关的冷却瞬变以及有利于计算机代码的开发和应用,用于加热与反应堆启动相关的播放瞬态。本文简要概述了本研究项目的当前结果。

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