MELCOR Computer Code Manuals

As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters.

Section: Discussionmentioning

confidence: 99%

Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept

Bruno

Paci

2015

“…Severe accident codes are usually used for this purpose. The codes can be divided into two categories that reflect the different level of details required from the analysis: (i) system analysis codes (e.g., MAAP [16], ASTEC [17], and MELCOR [1,2,[18][19][20][21]) that model progression of severe accidents and (ii) modeling tools that focus on specific components or phenomena (e.g., debris coolability, DECOSIM [14]; vessel lower head behavior, ANSYS/PECM [13]; and Zircaloy oxidation, MIDAC [22]) and require initial and boundary conditions provided from the system codes.…”

Section: Approachmentioning

confidence: 99%

“…The vessel breach condition was not implemented in the analysis, since the analysis is expected to provide the initial conditions for the analysis of vessel failure mode with ANSYS\PECM [13] and in-vessel debris coolability with DECOSIM code [14] within ROAAM+ framework for Nordic BWR [7,10]. We use MELCOR code versions 2.1 (rev7544) and 2.2 (rev9541) for prediction of the accident progression [1,2].…”

Section: Nordic Bwr Melcor Modelmentioning

confidence: 99%

“…Melt release characteristics provide initial conditions for the ex-vessel phase of SA progression, which involves phenomena that can threaten containment integrity (e.g., direct containment heating (DCH), fuel-coolant interactions (FCI), hydrogen combustion and detonation, molten-core-concrete interaction (MCCI), and ex-vessel debris formation and coolability). Assessment of severe accident progression is usually performed by means of SA analysis computer codes, such as MELCOR [1,2] and MAAP, and subject to aleatory uncertainty (e.g., different accident scenarios and human actions) and state-of-knowledge or epistemic uncertainty (models, modeling options, and sensitivity coefficients) [3][4][5].…”

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code

Galushin

Kudinov

2019

Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

“…To condense steam in the water, the model in MELCOR was used. MELCOR 2.1 has bubble physics model in the CVH/FL (Control volume hydrodynamics/Flow path) package [19]. It calculates pool scrubbing with RN (Radionuclide) package in MELCOR using SPARC90 model [20].…”

Section: Icrv-rdt Case Simulationmentioning

confidence: 99%

A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

Kim

Choi

Jeon

et al. 2018

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.