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

Uncertainty analysis of ATF Cr-coated-Zircaloy on BWR in-vessel accident progression during a station blackout

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The Use of MOX Fuel in the United States: Bibliography of Important Documents and Discussion of Key Issues

Severe accident analysis using dynamic accident progression event trees

In the current state of maturity of severe accident codes, the time has come to foster the systematic application of Best Estimate Plus Uncertainties (BEPU) in this domain. The overall objective of the HORIZON-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” is to quantify the uncertainties of severe accident codes (e.g., ASTEC, MAAP, MELCOR, and AC2) when modeling reactor and spent fuel pools accident scenarios of Gen II and Gen III reactor designs for the prediction of the radiological source term. To do so, different Uncertainty Quantification (UQ) methodologies are to be used for the uncertainty and sensitivity analysis. Innovative AM measures will be considered in performing these UQ analyses, in addition to initial/boundary conditions and model parameters, to assess their impact on the source term prediction. This paper synthesizes the major pillars and the overall structure of the MUSA project, as well as the expectations and the progress made over the first year and a half of operation.

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The EC MUSA project on management and uncertainty of severe accidents: Main pillars and status

An uncertainty analysis of the hydrogen source term for a station blackout accident in Sequoyah using MELCOR 1.8.5

As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents A series of 23 severe accident calculations was performed using MELCOR 185 for a four-loop Westinghouse reactor with an ice condenser containment The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465 Source term magnitude and timing results are presented consistent with the NUREG-1465 format For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups MELCORmore » results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465« less

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Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

This paper provides an overview of the U.S. Nuclear Regulatory Commission’s (NRC’s) project to develop a technical report summarizing the most important insights from its three State-of-the-Art Rea...

State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analyses for Station Blackout Scenarios

Phase 2 of the TMI-2 accident (core uncovery and melting) is revisited with the latest release of MELCOR (i.e., Version 1.8.5). An enhanced multi-ring multi-level hydro nodalization of the reactor core and upper plenum was developed to permit calculation of natural convection heat transfer between the core and upper internals. Uncertainties in boundary and initial conditions are investigated with particular attention given to the distribution of liquid in the RCS at the beginning of Phase 2 and to the histories of pressure and level in the steam generators. Special attention given to modeling the boiler sides of the once-through steam generators is presented. The modeling is designed to capture the rapid condensation of RCS vapor that would result from spraying cold auxiliary feedwater directly onto the upper portion of a voided steam generator tube bundle. Presentation of key Phase-2 accident signatures including fuel temperatures and hydrogen generation are presented. Comparisons are made between MELCOR calculations, TMI-2 data, and SCDAP/RELAP simulations. A largely improved MELCOR simulation of TMI-2 Phase 2 is obtained.Copyright © 2002 by ASME

Kenneth Charles Wagner

Papers

An uncertainty analysis of the hydrogen source term for a station blackout accident in Sequoyah using MELCOR 1.8.5

Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analyses for Station Blackout Scenarios

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

MELCOR integrated severe accident code application to safety assessment of high-temperature gas-cooled reactors