Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

Prošek, Andrej; Cizelj, Leon

doi:10.1155/2013/851987

Cited by 24 publications

(15 citation statements)

References 11 publications

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“…A problem of interest in the nuclear power industry involves the response of PWR pressure boundary components under SBO conditions [2]. SBO scenario involves a loss of off-site power, failure of the emergency diesel generators, failure of alternate current (AC) power and the eventual degradation of the reactor coolant pump (RCP) seals under the temperature effect and creep rupture in the surge line resulting in a long-term loss of coolant [3].…”

Section: Introductionmentioning

confidence: 99%

“…In the literature, there are many examples of stationblackout analysis, using severe accident codes to simulate, during the first 24 h, scenarios leading to RCS damage [4]. Vierow et al have studied [3,4] a TMLB hypothetical scenario of station blackout with a loss of AC power and no recovery of auxiliary feedwater to the steam generators at a 4-loop Westinghouse pressurized water reactor based on Zion. In this study, the authors used MELCOR, MAAP4 and SCDAP/RELAP5 severe accident codes.…”

Section: Introductionmentioning

confidence: 99%

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Numerical simulations of a creep rupture damage in Surry PWR primary circuit under TMLB accident

et al. 2020

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In the nuclear safety field, the thermomechanical phenomena that occur during an accident in a nuclear power plant (NPP) are of particular importance. During some reactor severe accidents, the decay energy of the core is transferred via the reactor cooling system (RCS) to other parts of this circuit. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities, i.e., the lower head vessel, the pressurizer surge line, the hot leg nozzles, etc. The potential for a primary circuit component rupture is of particular concern because fission products could be released into the environment by such a failure. The risk is higher in-vessel; the relocated debris toward the lower head vessel, following a potential melting of the core, leads to the rupture of the most important protective barrier. Primary circuit structures are subject to the effects of high pressure and high temperature throughout their service life. Such loadings generally lead to creep rupture. This risk is more likely when these structures experience transients or severe accidents. One of the most studied accidents is the station blackout (SBO). SBO without operator actions accident (TMLB' sequence) is considered as one of the most likely scenarios that may threaten the integrity of vulnerable RCS pressure boundaries. In this manuscript, we have evaluated, using the system code RELAP5/SCDAPSIM 3.4, the damage caused by rupture of some structures of the primary circuit of the Surry NPP. The structures analyzed are the surge line, the hot leg nozzles and the lower head vessel. RELAP5/SCDAPSIM 3.4 results indicate that surge line (hot leg) failures will be the first failures in the RCS pressure boundary.

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Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Numerical simulations of a creep rupture damage in Surry PWR primary circuit under TMLB accident

et al. 2020

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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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“…Several studies on the optimization of AM measures (e.g. primary-side and/or SG secondary-side depressurization) have been done for SBO scenarios through the long-term analyses of PWRs with best-estimate computer codes (Cherubini, et al, 2008;Tusheva, et al, 2012Tusheva, et al, , 2014Prošek and Cizelj, 2013). Some experimental data on PWR SBO scenarios with AM measures have been obtained by using such integral test facilities as the PSB-VVER in Russia (Bucalossi, et al, 2012) and the PKL in Germany (Umminger, et al, 2012).…”

Section: Introductionmentioning

confidence: 99%

ROSA/LSTF experiment on a PWR station blackout transient with accident management measures and RELAP5 analyses

Takeda

Ohtsu

2015

Mechanical Engineering Journal

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An experiment on a PWR station blackout transient with the TMLB' scenario and accident management (AM) measures was conducted using the rig of safety assessment/large scale test facility (ROSA/LSTF) at Japan Atomic Energy Agency under an assumption of non-condensable gas inflow to the primary system from accumulator (ACC) tanks. The TMLB' scenario involves prolonged complete loss of alternating current power and unavailability of turbine-driven auxiliary feedwater as well as malfunction of relief valves in primary system and steam generator (SG) secondary-side system. The AM measures considered in this study are SG secondary-side depressurization by fully opening the safety valves in both SGs with the start of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The LSTF test revealed that the primary pressure started to decrease when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after the completion of ACC coolant injection. The RELAP5 code well predicted the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of the SG U-tube collapsed liquid level and primary mass flow rate after the gas ingress. The SG coolant injection flow rate was found to significantly affect the peak cladding temperature and the ACC actuation time through the RELAP5 sensitivity analyses.

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Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

Cited by 24 publications

References 11 publications

Numerical simulations of a creep rupture damage in Surry PWR primary circuit under TMLB accident

Numerical simulations of a creep rupture damage in Surry PWR primary circuit under TMLB accident

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

ROSA/LSTF experiment on a PWR station blackout transient with accident management measures and RELAP5 analyses

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