Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

Dallman, R.J.; Gottula, R.C.; Holcomb, E.E.; Jouse, W.C.; Wagoner, S.R.; Wheatley, P.D.

doi:10.2172/6370654

Cited by 10 publications

(10 citation statements)

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“…Two other studies were included in the analysis that were performed by General Electric [33] and INEL [34]. The results of these two studies basically support the two previous studies.…”

Section: Iv-313supporting

confidence: 70%

Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

Kolaczkowski¹,

Lambright²,

Ferrell³

et al. 1986

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This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provide additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate.The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all AC power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors.The numerical results are largely driven by common mode failure probability estimates and to some extent, human error.Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis.iii/iv

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“…Two other studies were included in the analysis that were performed by General Electric [33] and INEL [34]. The results of these two studies basically support the two previous studies.…”

Section: Iv-313supporting

confidence: 70%

Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

Kolaczkowski¹,

Lambright²,

Ferrell³

et al. 1986

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Section: Cntrlvar 941mentioning

confidence: 99%

“…According to Table B-2 of NUREG/CR-4165 (Dallman, et al, 1987), the WW SP water level varies between 175.25 and 181.25 inches. According to Page 20 of NUREG/CR-4165 (Dallman, et al, 1987), the suction of the high-pressure injection (HPI) system pumps switches from the CST to the WW after the WW level reaches 181.25 inches. The water level in the WW at the end of the steady-state calculation is 4.5726 m. The switch to WW injection is modeled when the WW level has increased by 6 inches from the steady value or 4.5726m + 6 inches x 0.0254 (m/inch) = 4.7250 m, which is reflected in Trip 402.…”

Section: Cntrlvar 941mentioning

confidence: 99%

“…The water level in the WW at the end of the steady-state calculation is 4.5726 m. The switch to WW injection is modeled when the WW level has increased by 6 inches from the steady value or 4.5726m + 6 inches x 0.0254 (m/inch) = 4.7250 m, which is reflected in Trip 402. The bearings in the HPI pumps are assumed to fail when the fluid temperature in the SP reaches 361 K (190ºF) based on Page 20 of NUREG/CR-4165 (Dallman, et al, 1987). This value is reflected in Trip 502.…”

Section: Cntrlvar 941mentioning

confidence: 99%

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Risk-Informed Analysis for Enhanced Resilient Nuclear Power Plant with Initiatives including ATF, FLEX, and Advanced Battery Technology

Zhang

Hongbing

et al. 2021

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This report documents the activities performed by the Idaho National Laboratory (INL) during fiscal year (FY) 2021 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk-Informed Systems Analysis (RISA) Pathway as part of Enhanced Resilient Plant (ERP) Systems project. The purpose of the RISA Pathway research and development is to support plant owneroperator decisions with the aim to improve the economics, reliability, and maintain the high levels of safety of current nuclear power plants over periods of extended plant operations. The concept of ERP refers to a nuclear power plant (NPP) that is enhanced with various industry initiatives such as accidenttolerant fuel (ATF), optimal use of diverse and flexible coping strategy (FLEX), enhancements to plant components and systems, the incorporation of augmented or new passive cooling systems, and advanced battery technology with extended capacity. The objective of the ERP research effort is to use the RISA methods and toolkit in industry applications, including methods development and early demonstration of technologies, in order to enhance existing reactors' safety features and to substantially reduce operating costs through risk-informed approaches to plant design modifications and their characterization.

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“…In order to determine the consequences of these types of sequences, calculations were done with the RELAP5/MOD2 code [23] with a model of the Browns Ferry plant based on previous studies [24]. The reactor was assumed to be initially operating at nominal power.…”

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confidence: 99%

Reactivity accidents: A reassessment of the design-basis events

Diamond¹,

Hsu²,

Fitzpatrick³

et al. 1990

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This paper summarizes a study of light water reactor event sequences which have been investigated for their potential to result in reactivity accidents with severe consequences.The study is an outgrowth of the concern which arose after the accident at Chernobyl and was recommended by the report of the U.S. Nuclear Regulatory Commission (NRC) on the implications of that accident (NUREG-1251). The work was done for the NRC to reconfirm or bring into question previous judgements on reactivity events which must be analyzed for licensing. Event sequences were defined and then a probabilistic assessment was completed to estimate the frequency of the reactivity event and/or a deterministic calculation was completed to estimate the consequences to the fuel. Using the results of this analysis, analysis done by others, and a set of screening criteria developed within this study, judgements were made for each sequence as to its importance, and recommendations were made as to whether the NRC ought to be considering the important sequences as part of the design basis or for further, more detailed, investigation. SummaryThe required safety analysis for light water reactors includes various reactivity events which must result in acceptable consequences in order for plants to be licensed. Although it is possible to hypothesize reactivity events that might lead to unacceptable consequences, in the past these unanalyzed events were either precluded by specific design features or required combinations of causative, events which were judged too improbable to be of concern. The objective of this study was to analyze these latter events, and events newly identified in this study, to reconfirm or bring into question previous judgements on the adequacy of the events selected as the design-basis, and to determine if any events needed more analysis in the context of severe accident research. This was done by performing probabilistic and/or deterministic calculations and, where available, using the research of others.

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Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

Cited by 10 publications

References 0 publications

Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

Risk-Informed Analysis for Enhanced Resilient Nuclear Power Plant with Initiatives including ATF, FLEX, and Advanced Battery Technology

Reactivity accidents: A reassessment of the design-basis events

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