Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

Rüger, Carlo; Luckas, W.J.

doi:10.2172/6428071

Cited by 7 publications

(9 citation statements)

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“…The leak simulation was based on the expected behavior of Westinghouse RCP seals, which are used in the Surry plant. 25 Specifically, Westinghouse seals are expected to leak 21 gpm of subcooled water per RCP at nominal operating conditions if seal cooling water is lost. 26 Subsequent seal degradation and failure is expected as a TMLB' accident progresses because the seals are not designed for exposure to high temperature steam.…”

Section: Surry Analyses With Rcp Seal Leaksmentioning

confidence: 99%

“…The simulation was based on the expected behavior of Byron Jackson RCP seals, which are used in the ANO2 plant. 25 Specifically, Byron Jackson pump seals are expected to leak 1.5 gpm of subcooled water per RCP at nominal operating conditions if seal cooling water is lost and an estimated maximum of 220 gpm per RCP. 29 Based on that information, a seal leak area equivalent to 1.5 gpm per RCP was introduced in Cases ANO2-03 and ANO2-04 at TMLB' accident initiation since seal cooling water is lost when AC power is lost.…”

Section: Ano2 Analyses With Rcp Seal Leaksmentioning

confidence: 99%

“…Surry and Zion plants, like all other Westinghouse PWRs, use Westinghouse RCPs. 25 Westinghouse pump seals are expected to leak 21 gpm of subcooled water per RCP at nominal operating conditions if INEEL/EXT-98-00286, Rev. 1 seal cooling water is lost.…”

Section: Rcs Depressurization With Rcp Seal Leaksmentioning

confidence: 99%

“…The Calvert Cliffs plant, like all Combustion Engineering PWRs other than the Palo Verde plants, uses Byron Jackson RCPs. 25 (Klein, Schanzlin, & Becker RCPs are used in Palo Verde 1, 2, and 3. 25 ) Byron Jackson pump seals are expected to leak 1.5 gpm of subcooled water per RCP at nominal operating conditions if seal cooling water is lost and an estimated maximum of 220 gpm per RCP.…”

Section: Rcs Depressurization With Rcp Seal Leaksmentioning

confidence: 99%

“…25 (Klein, Schanzlin, & Becker RCPs are used in Palo Verde 1, 2, and 3. 25 ) Byron Jackson pump seals are expected to leak 1.5 gpm of subcooled water per RCP at nominal operating conditions if seal cooling water is lost and an estimated maximum of 220 gpm per RCP. 29 Based on that information, a decision was made to evaluate seal leaks of 110 and 220 gpm per RCP.…”

Section: Rcs Depressurization With Rcp Seal Leaksmentioning

confidence: 99%

See 4 more Smart Citations

SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

Knudson¹,

Ghan²,

Dobbe³

1998

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Section: Surry Analyses With Rcp Seal Leaksmentioning

confidence: 99%

Section: Ano2 Analyses With Rcp Seal Leaksmentioning

confidence: 99%

Section: Rcs Depressurization With Rcp Seal Leaksmentioning

confidence: 99%

Section: Rcs Depressurization With Rcp Seal Leaksmentioning

confidence: 99%

Section: Rcs Depressurization With Rcp Seal Leaksmentioning

confidence: 99%

See 3 more Smart Citations

SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

Knudson¹,

Ghan²,

Dobbe³

1998

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Safety margin sensitivity analysis for model selection in nuclear power plant probabilistic safety assessment

Maio

Picoco

Zio

et al. 2017

Reliability Engineering & System Safety

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The safety assessment of Nuclear Power Plants makes use of Thermal-Hydraulic codes for the quantification of the safety margins with respect to upper/lower safety thresholds, when postulated accidental scenarios occur. To explicitly treat uncertainties in the safety margins estimates within the Risk-Informed Safety Margin Characterization (RISMC) framework, we resort to the concept of Dynamic Probabilistic Safety Margin (DPSM). We propose to add to the framework a sensitivity analysis that calculates how much the Thermal-Hydraulic (TH) code inputs affect the DPSM, in support to the selection of the most proper probabilistic safety assessment method to be used for the problem at hand, between static or dynamic methods (e.g., Event Trees (ETs) or Dynamic ETs (DETs), respectively). Two case studies are considered: firstly a Station Black Out followed by a Seal Loss Of Coolant Accident (LOCA) for a 3-loops Pressurized Water Reactor (PWR), whose dynamics is simulated by a MAAP5 model and, secondly, the accidental scenarios that can occur in a U-Tube Steam Generator, whose dynamics is simulated by a SIMULINK model. The results show that the sensitivity analysis performed on the DPSM points out that an ET-based analysis is sufficient in one case, whereas a DET-based analysis is needed for the other case.

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Quantification of risk dilution induced by correlation parameters in dynamic probabilistic risk assessment of nuclear power plants

Kubo

Tanaka

Ishikawa

2023

Proceedings of the Institution of Mechanical Engineers, Part O:

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Nuclear power plants are critical infrastructures that produce electricity. However, accidents in nuclear power plants can cause considerable consequences such as the release of radioactive materials. Therefore, appropriately managing their risk is necessary. Various nuclear regulatory agencies employ probabilistic risk assessment (PRA) to effectively evaluate risks in nuclear power plants. Dynamic PRA has gained popularity because it allows for more realistic assessment by reducing the assumptions and engineering judgments related to time-dependent failure probability and/or human-action reliability in the conventional PRA methodology. However, removing all assumptions and engineering judgments is difficult; thus, the risk analyst, for example, the regulator, must understand their effects on the assessment results. This study focuses on “risk dilution,” which emerges from the assumptions about uncertainty. Dynamic PRA of a station blackout sequence in a boiling-water reactor was performed using the dynamic PRA tool, namely, Risk Assessment with Plant Interactive Dynamics (RAPID) and the severe-accident code Thermal–Hydraulic Analysis of Loss of Coolant, Emergency Core Cooling, and Severe Core Damage version 2 (THALES2), which altered the correlation parameters among the uncertainties of the events that occurred in sequence. The results demonstrated that the conditional core-damage probability and mean value of the core-damage time varied from 0.27 to 0.47 and from 7.1 to 8.7 h, respectively. When the dynamic PRA results are used for risk-informed decision making, the decision maker should adequately consider the effect of risk dilution.

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Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

Cited by 7 publications

References 0 publications

SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

Safety margin sensitivity analysis for model selection in nuclear power plant probabilistic safety assessment

Quantification of risk dilution induced by correlation parameters in dynamic probabilistic risk assessment of nuclear power plants

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