MELCOR Computer Code Manuals Volume 1: Primer and Users' Guide

Humphries, Larry L.; Figueroa, Victor G.; Young, Michael; Louie, David; Reynolds, John J.

doi:10.2172/1433069

Cited by 26 publications

(32 citation statements)

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“…This package includes also a specific model to simulate the Iodine behavior in gaseous and liquid phases. Specific information on the different models implemented in each MELCOR module can be found in the code manuals [21].…”

Section: Codes Employedmentioning

confidence: 99%

“…The 7 FLPs connect the different CVs, while the 8 HSs simulate the sump outer wall (WSU), the semi-elliptic bottom, and top of the vessel (WB1, W2T, and W4T), the cylindrical outer wall (W3, W4), and the three condensers (WET and DRY). The four remaining HSs were utilized to provide an aerosol settling structure on the bottom of the C1, C2, C3, and C4 CVs as suggested in the MELCOR User's Guide [21]. Except WET and DRY, all these HSs were simulated with an imposed outer temperature evolving in time according to [9], with an inner temperature based on the heat exchange with their surrounding CVs.…”

Section: Employed Nodalizations and Boundary Conditionsmentioning

confidence: 99%

“…In both codes, additional HSs on the bottom of C * CVs are added to provide a settling surface as suggested in [21]. Such HSs have no influence on the MELCOR calculations, but in ASTEC they improve the overall aerosol results.…”

Section: Characteristic Length Relationsmentioning

confidence: 99%

“…From the performed analysis, a CL value of few cm was deemed necessary to better reproduce the thermal-hydraulic transient of the Phébus FPT-0 test. As stated in the MELCOR manual [21], the heat transfer regime is defined based on the ratio between the Reynolds (Re) and the Grashof (Gr) numbers. When the ratio Re 2 /Gr is below 1, a natural convection regime is considered, and if above 10, a forced convection regime is established, and an intermediate condition is considered in between.…”

Section: Thermal-hydraulics Sensitivity Analysesmentioning

confidence: 99%

“…This result is not due to a user's assumption, but it is the normal resulting behavior of the water-film tracking model, as explicitly declared by the sensitivity coefficient no. 7136 (solubility of RN classes in water films) [21]. Such value could be changed by the user, but no sensitivity analyses have been performed due to the absence of information about the repartition coefficient between the liquid film and the structure.…”

Section: Mel Be Expmentioning

confidence: 99%

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Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

Bruno

Paci

2017

Science and Technology of Nuclear Installations

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The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

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Section: Codes Employedmentioning

confidence: 99%

Section: Employed Nodalizations and Boundary Conditionsmentioning

confidence: 99%

Section: Characteristic Length Relationsmentioning

confidence: 99%

Section: Thermal-hydraulics Sensitivity Analysesmentioning

confidence: 99%

Section: Mel Be Expmentioning

confidence: 99%

See 3 more Smart Citations

Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

Bruno

Paci

2017

Science and Technology of Nuclear Installations

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Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Zhao

Zou²,

et al. 2022

Intl J of Energy Research

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Summary A simulation calculation model in the case of a severe accident such as a loss of coolant accident (LOCA) was established and calculated by MELCOR code in a marine pressurized water reactor. The effects of severe accidents on the hydrogen source term, release of the radionuclide source term, and the accident process were investigated, and the radiation consequences were analyzed. LOCA sensitivity analysis was conducted with different break sizes; the release and diffusion of source terms were analyzed. The results indicated that the hydrogen yield in the reactor core was dependent on the temperature, and on the residual water in the reactor core, which has no direct relationship with the break size. The break size directly affected the accident progression. However, when the break exceeded a certain size, it did not affect the LOCA process. The break size may not change to change the release amount and migration rule of radionuclides such as Xe and I, and the fission product CsI, as final containment is in dynamic equilibrium in LOCA conditions. The contents of fission products released into the containment vessel, major loop, cavity, and the environment are sensitive to the break size. However, no special correlation is observed between the amount released and the break size. Highlights The hydrogen yield in the reactor core was dependent on the temperature, and on the residual water in the reactor core, which has no direct relationship with the break size. The break size may not change to change the release amount and migration rule of radionuclides such as Xe, I, and CsI, as final containment is in dynamic equilibrium in LOCA conditions. No particular correlation is observed between the amount of radionuclide released and the break size.

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Investigation on influence of crust formation on VULCANO VE-U7 corium spreading with MPS method

Yasumura

Yamaji

Furuya

et al. 2017

Annals of Nuclear Energy

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MELCOR Computer Code Manuals Volume 1: Primer and Users' Guide

Cited by 26 publications

References 0 publications

Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Investigation on influence of crust formation on VULCANO VE-U7 corium spreading with MPS method

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