Simulation of SB-LOCA of typical PWR with MELCOR code

Malicki, Mateusz; Pieńkowski, L.; Skolik, Katarzyna

doi:10.1088/1755-1315/214/1/012071

Cited by 3 publications

(3 citation statements)

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“…The general result showed a good agreement between the reference data and the MELCOR calculation (i.e. considering the results of the consulted literature [11][12][13][14] ) being used as the basis for this work.…”

Section: Model For the Reference Plant Reactor Cooling Systemsupporting

confidence: 60%

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Modeling Lofw in a PWR Using Melcor

et al. 2021

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The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity.

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Section: Model For the Reference Plant Reactor Cooling Systemsupporting

confidence: 60%

“…Peak fuel cladding temperature (Figure 10) 260. 11 Maximum hydrogen production (0.0987 Kg) (Figure 11) 242. 26 Maximum amount of hydrogen in core (0.003 Kg) (Figure 11) 300.…”

Section: Sequence Of Eventsmentioning

confidence: 99%

Modeling Lofw in a PWR Using Melcor

et al. 2021

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“…Considering the consulted literature [17][18][19][20], the comparison based in relative error shows a good agreement between the reference values presented in the RELAP5/MOD2 model and MELCOR calculationi.e., the evaluated error is in the same magnitude order as the consulted papers [21] -, confirming that the MELCOR input model represents with good accuracy the reference plant.…”

Section: Results For the Steady State Conditionmentioning

confidence: 57%

MELCOR steady state calculation of the generic PWR of 40MWth

2021

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After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

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Study of the DVI-LOCA in the AP1000-like reactor with MELCOR code

Włostowski,

Darnowski

2024

Annals of Nuclear Energy

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Simulation of SB-LOCA of typical PWR with MELCOR code

Cited by 3 publications

References 1 publication

Modeling Lofw in a PWR Using Melcor

Modeling Lofw in a PWR Using Melcor

MELCOR steady state calculation of the generic PWR of 40MWth

Study of the DVI-LOCA in the AP1000-like reactor with MELCOR code

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