MELCOR Code Change History (Revision 14959 to 18019)

Humphries, Larry L.; Phillips, Jesse; Schmidt, Rodney C.; Beeny, Bradley; Louie, David; Bixler, Nathan E.

doi:10.2172/1756139

Cited by 7 publications

(19 citation statements)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…FCELR and FCELA factors account for the radiative exchange among cells in the "core" region: more precisely, they represent the exchanges occurring radially outward and axially upward, respectively. FCELR and FCELA are "effective" view factors: the default value (0.1) takes into account the fact that, when considering surfaces far from the cell boundary, the line of sight is intercepted by other surfaces, greatly reducing the differential view factors [32]. Values higher than the default one are investigated to evaluate potential effects on the system, also considering that the HI-STORM "core-like" region is smaller than that of a LWR, for which the default value has been derived.…”

Section: Uncertainty Configurationmentioning

confidence: 99%

See 1 more Smart Citation

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

et al. 2022

View full text Add to dashboard Cite

Nuclear safety relies to a good extent on thoroughly validated codes. However, code predictions are affected by uncertainties that need to be quantified for a more accurate evaluation of safety margins. In this regard, the present paper proposes a preliminary uncertainty and sensitivity analysis of the thermal behavior of a concrete-based dry cask for spent nuclear fuel storage, employing the MELCOR code and a series of MATLAB scripts. As thermal behavior is of utmost importance for the fulfillment of United States Nuclear Regulatory Commission (USNRC) safety requirements, the Peak Cladding Temperature (PCT) has been addressed as the key Figure of Merit (FOM). Variables related to the main heat transfer mechanisms have been selected as input parameters for the uncertainty quantification, whereas heat source and heat sink, namely decay power and external air temperature, have been dealt with in a separate sensitivity analysis. The results show that the selected parameters have a weak influence on the PCT, whereas it is strongly related to the decay power and external air temperature values. In any case, PCT stays below the regulatory threshold even under the considered off-normal conditions.

show abstract

Section: Uncertainty Configurationmentioning

confidence: 99%

“…The analysis focuses on the PCT, being the variable under regulatory surveillance, and on the uncertainties embedded in the values predicted by the code. A reasonable accuracy and a significant reduction of the computational cost make the MELCOR code [31,32] suitable for performing UQ in an easier and faster way.…”

Section: Introductionmentioning

confidence: 99%

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

et al. 2022

View full text Add to dashboard Cite

show abstract

“…In this work, the MELCOR 2.0 Zion LBLOCA input deck was modified for the newer MELCOR Version 2.2, Build 20234. The main relevant differences between MELCOR 2.0 and 2.2 are the corrections to the eutectics model method, the addition of a user-defined material capability, and improvements to the modeling of ATFs with extended options for oxidation definition [43]. The following subsections describe the modeling details.…”

Section: Control Rod Materials Ag-in-cdmentioning

confidence: 99%

“…This study, therefore, used MELCOR 2.2. MELCOR 2.2 contains the new capability to create User-Defined Materials (UDMs) that allow the explicit definition of additional, non-default materials in MELCOR [47] and [48]. This new capability is used in this work to define the FeCrAl cladding material and FeCrAl oxide cladding material.…”

Section: Fecral and Fecral Oxide Definitionmentioning

confidence: 99%

Safety Analysis for Accident Tolerant Fuels with Increased Enrichment and Extended Burnup

Choi

Ogujiuba

Rollins

et al. 2022

View full text Add to dashboard Cite

“…Likewise, advection of radionuclides between control volumes is done using CVs flows and wash-off of radionuclides deposited on heat structures, is determined from the drainage of water films calculated by the HS package. The RN package determines decay heat power for current radionuclide inventories from the Decay Heat (DCH) package when requested by each of these packages [21].…”

Section: Melcor Code Descriptionmentioning

confidence: 99%

Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

et al. 2021

View full text Add to dashboard Cite

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

show abstract

MELCOR Code Change History (Revision 14959 to 18019)

Cited by 7 publications

References 0 publications

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

Safety Analysis for Accident Tolerant Fuels with Increased Enrichment and Extended Burnup

Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

Contact Info

Product

Resources

About