MELCOR 1.8.5 Simulation of TMI-2 Phase 2 With an Enhanced 2-Dimensional In-Vessel Natural Circulation Model

Gauntt, Randall O.; Ross, Kyle; Wagner, Kenneth Charles

doi:10.1115/icone10-22321

Cited by 5 publications

(4 citation statements)

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“…Severe accident code simulations for the TMI-2 scenario have been carried out with SCDAP/RELAP5 (Hohorst et al, 1994), MELCOR (Gauntt et al, 2002) and MAAP4 (Paik et al, 1995). Comparing the different severe accident codes' predictions against plant data is an essential test of the codes' accuracy and will provide additional information on the relative merits of the various severe accident models.…”

Section: Discussionmentioning

confidence: 99%

Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes

Vierow¹,

Liao²,

Johnson³

et al. 2004

Nuclear Engineering and Design

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Based on knowledge obtained from experimental programs, considerable progress has been made in severe accident code development. The three leading severe accident codes used in the U.S., MELCOR, MAAP4 and SCDAP/RELAP5, are compared herein as part of an evaluation of the relative state of severe accident modeling in each of the codes.The MELCOR code is evolving from a probabilistic risk assessment tool to a best-estimate severe accident system analysis code. Some advantages that MELCOR has are the capability to evaluate containment behavior and the source term to the environment, and the great modeling flexibility that the control volume approach and control functions afford. The MAAP4 code was developed to perform fast-running full simulations of severe accidents. Due in part to the simplified form of the conservation equations and the coarser discretization of the reactor systems, MAAP4 has calculation times far shorter than those of the other codes while producing credible results. The SCDAP/RELAP5 code contains more mechanistic physics models than the other codes for both severe accident and thermal-hydraulic phenomena and has undergone extensive validation against plant and experimental data.The codes' overall attributes, relevant physics models and calculation results are compared herein. A hypothetical TMLB' scenario (station blackout with no recovery of auxiliary feedwater) at a 4-loop pressurized water reactor (PWR) was calculated by all three codes. Detailed plots showed that, despite considerable differences in the codes themselves, the calculation results of the codes are very similar in terms of thermal-hydraulic and core degradation response. There are minor discrepancies in various timings of phenomena, which are within the uncertainties of the code numerical computation and the physics models. The thermal challenge to the primary loop radionuclide barrier, such as the steam generator tube heat structures, is also compared among the three codes.

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Section: Discussionmentioning

confidence: 99%

Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes

Vierow¹,

Liao²,

Johnson³

et al. 2004

Nuclear Engineering and Design

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“…A modified version of the MELCOR 1.8.5 code [15,16] was used to analyse the ex-vessel TBM helium coolant leaks accident scenario. The input model developed for MELCOR includes the TBM, the TBM ancillary equipment and ITER relevant enclosures.…”

Section: Methods Of Analysismentioning

confidence: 99%

Safety analysis of the US dual coolant liquid lead–lithium ITER test blanket module

Merrill¹,

Reyes²,

Sawan³

et al. 2007

Nucl. Fusion

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The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

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“…The TMI-2 accident was caused by a small-break LOCA in a two-loop Babcock and Wilcox (B&W) pressurized water reactor (PWR). The MELCOR input model used for this accident analysis is that developed by SNL-NM and Innovative Technology Solutions Corporation, Albuquerque, NM [3].…”

Section: Initial Melcor Modifications: Materials Substitutionmentioning

confidence: 99%

“…As mentioned previously, the MELCOR model used the Zircaloy cladding model developed by SNL-NM and Innovative Technology Solutions Corporation, Albuquerque, NM, as the base case for this study [3]. This model was used as received from SNL-NM.…”

Section: Sic Performance During the Tmi Eventmentioning

confidence: 99%

Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

Merrill

Bragg-Sitton

Humrickhouse

2017

Nuclear Engineering and Design

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A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation's nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry's success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO 2-Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications made to a version of the MELCOR code by the Idaho National Laboratory (INL) that allows its application to alternate fuel-cladding systems. Preliminary analysis results are presented in this paper for candidate silicon carbide and FeCrAl cladding concepts.

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MELCOR 1.8.5 Simulation of TMI-2 Phase 2 With an Enhanced 2-Dimensional In-Vessel Natural Circulation Model

Cited by 5 publications

References 0 publications

Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes

Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes

Safety analysis of the US dual coolant liquid lead–lithium ITER test blanket module

Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

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