Station blackout mitigation strategies analysis for Maanshan PWR plant using TRACE

Lin, Hao-Tzu; Wang, Jong-Rong; Huang, Kuang-Tzu; Shih, Chunkuan; Chiang, Show-Chyuan; Liu, Chia-Chuan

doi:10.1016/j.anucene.2015.11.015

Cited by 9 publications

(2 citation statements)

References 16 publications

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“…In recent years, stress tests have been carried out worldwide to study the effects and preparedness of the nuclear installations against SBO sequences (Queral et al, 2016;Tatu and Kim, 2017). Efforts have been carried out in several countries to improve the safety systems involved in such sequences (Lin et al, 2016). For instance, in Spain, emergency mobile pumps have been prepared to be able to inject water to the secondary systems (París et al, 2019).…”

Section: Introductionmentioning

confidence: 99%

Modelling guidelines for safety analysis of Station Black Out sequences based on experiments at the PKL test facility

Freixa

Martinez-Quiroga

Reventós

2020

Annals of Nuclear Energy

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Section: Introductionmentioning

confidence: 99%

Modelling guidelines for safety analysis of Station Black Out sequences based on experiments at the PKL test facility

Freixa

Martinez-Quiroga

Reventós

2020

Annals of Nuclear Energy

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“…Steam generator (SG) secondary-side depressurization through the SG valve(s) is one of major AM measures to cool and depressurize the primary system via natural circulation (NC) because of steam condensation in the SG U-tubes especially when high-pressure injection system of emergency core cooling system (ECCS) is totally failed during accidents and transients in a pressurized water reactor (PWR). Several studies [2][3][4][5][6] have been conducted to investigate thermal hydraulic responses in SBO scenarios with AM measures of PWRs by calculations with best-estimate computer codes. Some researchers [7,8] have analyzed loss of primary coolant and SBO events in PWRs employing severe accident computer codes, but under the condition modeling for the pressure vessel internals except the core and the downcomer was more simplified than that in the best-estimate computer codes.…”

Section: Introductionmentioning

confidence: 99%

ROSA/LSTF Tests and Posttest Analyses by RELAP5 Code for Accident Management Measures during PWR Station Blackout Transient with Loss of Primary Coolant and Gas Inflow

Takeda

Ohtsu

2018

Science and Technology of Nuclear Installations

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Three tests were carried out with the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated accident management (AM) measures during station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total failure of high-pressure injection system in a pressurized water reactor. As the AM measures, steam generator (SG) secondary-side depressurization was done by fully opening the relief valves in both SGs, and auxiliary feedwater was injected into the secondary-side of both SGs simultaneously. Conditions for the break size and the onset timing of the AM measures were different among the three LSTF tests. In the three LSTF tests, the primary pressure decreased to a certain low pressure of below 1 MPa with or without the primary depressurization by fully opening the relief valve in a pressurizer as an optional AM measure, while no core uncovery took place through the whole transient. Nonuniform flow behaviors were observed in the SG U-tubes under natural circulation (NC) with nitrogen gas depending probably on the gas accumulation rate in the two LSTF tests that the gas accumulated remarkably. The RELAP5/MOD3.3 code predicted most of the overall trends of the major thermal hydraulic responses observed in the three LSTF tests. The code, however, indicated remaining problems in the predictions of the primary pressure, the SG U-tube collapsed liquid levels, and the NC mass flow rate after the nitrogen gas ingress as well as the accumulator flow rate through the analyses for the two LSTF tests, where the remarkable gas accumulation occurred.

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Assessment of fuel-rod meltdown in a severe accident at Bushehr nuclear power plant (BNPP)

2019

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Station blackout mitigation strategies analysis for Maanshan PWR plant using TRACE

Cited by 9 publications

References 16 publications

Modelling guidelines for safety analysis of Station Black Out sequences based on experiments at the PKL test facility

Modelling guidelines for safety analysis of Station Black Out sequences based on experiments at the PKL test facility

ROSA/LSTF Tests and Posttest Analyses by RELAP5 Code for Accident Management Measures during PWR Station Blackout Transient with Loss of Primary Coolant and Gas Inflow

Assessment of fuel-rod meltdown in a severe accident at Bushehr nuclear power plant (BNPP)

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