MELCOR Model Research Articles

Gas-cooled reactor thermal hydraulic analyses with MELCOR

Pursuant to the Energy Policy Act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the reference design for the Next Generation Nuclear Plant (NGNP). Stemming from a U.S. Nuclear Regulatory Commission (NRC) HTGR research initiative, a need was identified for validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform licensing analyses. Because the NRC has used MELCOR for light water reactors (LWR) in the past and because MELCOR was recently updated to include gas-cooled reactor (GCR) physics models, MELCOR is among the system codes of interest to the NRC. This paper describes MELCOR modeling of the General Atomics' Modular High Temperature Gas-Cooled Reactor (MHTGR). The MHGTR is a suitable design for demonstration of MELCOR GCR modeling competency for two reasons: 1) the MHTGR is a predecessor to the more advanced General Atomics’ Gas-Turbine Modular High Temperature Reactor (GTMHR), and 2) experimental data useful for benchmark calculations may soon become available. Using the most complete literature references available for the MHTGR design, researchers at Texas A&M University (TAMU) constructed a MELCOR input deck for the MHTGR to partially validate MELCOR GCR modeling capabilities. Normal and off-normal system operating conditions were modeled with appropriate boundary and initial conditions. MELCOR predictions of system response were obtained for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) scenarios. Code results were checked against nominal MHTGR design parameters, physical intuition, and anticipated GCR thermal hydraulic response. No inherent deficiencies in MELCOR modeling capability were observed, suggesting that the newly-implemented GCR models are adequate for systems-level analysis. If and when experimental benchmark data becomes available, further validation activities may proceed given the modeling efforts discussed herein.

A MELCOR model of Fukushima Daiichi Unit 1 accident

A MELCOR model of Fukushima Daiichi Unit 1 accident was developed. The model is based on publicly available information, and the MELCOR input file is published as Electronic Supplementary data with this paper. In order to reproduce the measured containment pressure, it was necessary to model a leak from the reactor coolant system. Recirculation pump seal leak, starting 5h after the earthquake, was assumed in this study. The reactor pressure vessel lower head was calculated to fail, and all fuel was discharged to the containment. Almost all of the radioactive noble gases and about 0.05% of the cesium inventory were released to the environment, according to this calculation. Calculated release rates from Units 1, 2, and 3 were compared with measured radiation dose rate in the plant area.

A MELCOR model of Fukushima Daiichi Unit 3 accident

A MELCOR model of the Fukushima Daiichi Unit 3 accident was developed. The model is based on publicly available information, and the MELCOR input file is published as electronic supplementary data with this paper. According to the calculation, the reactor pressure vessel lower head failed about 53h after the earthquake. At the end of the calculation, 30% of the fuel was still inside the reactor and 70% had been discharged to the containment. Almost all of the radioactive noble gases and 0.95% of the cesium inventory were released to the environment during the accident.

MELCOR Accident Analysis for ARIES-ACT

We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

MELCOR model of divertor cooling loop and divertor ex-vessel LOCA analysis for the ITER plant

A MELCOR input model of the divertor components and associated heat transfer system has been developed in order to analyse the divertor ex-vessel LOCA that potentially could result in radioactive releases to the environment. This paper presents the results of some parameter studies related to the DV ex-vessel LOCA: (i) three different in-vessel break sizes: one toroidal ring break with a size of 0.32 m 2, one-half ring break of 0.16 m 2 and one-quarter ring break of 0.08 m 2. (ii) Two different ex-vessel break locations were used: before heat exchanger and after heat exchanger but before pump.