MODELING LOFW IN A PWR USING MELCOR
DOI:
https://doi.org/10.15392/bjrs.v8i3A.1355Keywords:
severe accident, MELCOR, PSA.Abstract
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.
Downloads
References
U.S. Nuclear Regulatory Commission –USNRC. Probabilistic Risk Assessment and Severe accident Evaluation for New Reactors, NUREG-0800, Section 19.0, Washington, D. C., (2007).
M. R. Hayns. The Evolution of Probabilistic Risk Assessment in the Nuclear Industry, Institution of Chemical Engineers, Trans I, v. 77, Part B, (1999).
M. R. Martins, P. F. F. Melo and M. C. Maturana. Methodology for system reliabilityanalysis during the conceptual phase of complex system design considering humanfactors”. In: Proceeding of the ANS PSA 2015 International Topical Meeting onProbabilistic Safety Assessment and Analysis, Sun Valley, ID, April 26-30, 2015, onCD-ROM, American Nuclear Society, LaGrange Park, IL, (2015).
M. C. Maturana, L. L. Bruno, and M. R. Martins, Application of Fire PSA inDefining System Reliability Criteria: Detection and Suppression Systems in I&CElectrical Panel Room, In: Proceedings of Probabilistic Safety Assessment andManagement – PSAM 14, Los Angeles, CA, September, (2018).
L.L. Humphries, et al. MELCOR Computer Code Manuals Vol. 1: Primer andUsers’ Guide Version 2.2.9541 (2017).
L.L. Humphries, et al. MELCOR Computer Code Manuals Vol. 2: Primer andUsers’ Guide Version 2.2.9541 (2017).
N. N. Araújo, M. C. Maturana, M. R. Gual, Analytical Simulation of a PWR UsingMELCOR Code, In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, ,Santos, SP, Brazil, October 21-25 (2019).
J. Cardoni, R. Gauntt, D. Kalinich and J. Phillips, MELCOR simulations of severeaccident at Fukushima Daiichi Unit 3, Nuclear Technology, 2, Volume 186 (2014).
T. Sevon, A MELCOR model of Fukushima Daiichi Unit 3 accident, NuclearEngineering and Design, 284, pp. 80-90, April (2015).
M. Genta Maragni, A. Belchior Junior and J. A. Onoda Pessanha. Modelagem eestado estacionário do reator da INAP com o RELAP5/MOD2, In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, (1997).
M. Malcki, L. Pienkowski, K. Skolik, Simulation of SB-LOCA of typical PWR withMELCOR code, In: IOP Conference Series Earth and Environmental Science,214:012071, January (2019).
M Pescarini, F. Mascari, D. Mostacci and F De Rosa, Analysis of unmitigated largebreak loss of coolant accidents using MELCOR code, Journal of Physics ConferenceSeries 923, Volume 1,012009, November (2017).
Y. Jin, W. Xu, X. Liu, In- and ex-vessel coupled analysis of IVR-ERVCphenomenon for large scale PWR, Annals of Nuclear Energy, Volume 80, pp. 322-337, June (2015).
M. Pavlova, P. P Groudev and V. Hadjiev, Development and validation of VVER-1000 input deck for severe accident calculations with MELCOR Computer Code, In: XVInternational School on Nuclear Physics, Neutron Physics and NuclearEnergy, Varna, Bulgaria, October (2003).
NUREG‐0800: Standard Review Plan for the Review of Safety Reposts for NuclearPower Plants: LWR Edition, Revision 2, June (2007).
USNRC Regulations Title 10, Code of Federal Regulations (CRF), Part 50 - DomesticLicensing of Production and Utilization Facilities, Section 50.44 Combustible gascontrol for nuclear power reactors, January 1( 2011).
Y. Abou-Rjeily, G. Cénérino, A. Drozd, S. Lee, J. Misak, C.O. Park, et al.International Atomic Energy Agency, Mitigation of Hydrogen Hazards in SevereAccident in Nuclear Power Plants, IAEA-TECDOC-1661 (2011).
J. Yanez J, M. Kuznetso, A. Souto‐Iglesias A. An analysis of the hydrogen explosionin the Fukushima‐Daiichi accident, Int J Hydrogen Energy, Volume 40, pp. 8261‐8280 (2015).
AptPlot 6.8.0 / ACS Plug-in 2.3.0 - released December 1 (2017).
Published
Issue
Section
License
Copyright (c) 2021 Brazilian Journal of Radiation Sciences
This work is licensed under a Creative Commons Attribution 4.0 International License.
Licensing: The BJRS articles are licensed under a Creative Commons Attribution 4.0 International License, which permits use, sharing, adaptation, distribution and reproduction in any medium or format, as long as you give appropriate credit to the original author(s) and the source, provide a link to the Creative Commons license, and indicate if changes were made. The images or other third party material in this article are included in the article’s Creative Commons license, unless indicated otherwise in a credit line to the material. If material is not included in the article’s Creative Commons license and your intended use is not permitted by statutory regulation or exceeds the permitted use, you will need to obtain permission directly from the copyright holder. To view a copy of this license, visit http://creativecommons.org/licenses/by/4.0/
How to Cite
