Uncertainty analyses using the Melcor severe accident analysis code
Creators
- 1. Sandia National Laboratories, Analysis and Modeling Dept., Albuquerque NM (United States)
Description
The MELCOR code is a detailed system level computer code capable of performing integrated self-consistent analyses of severe accident progression in commercial nuclear power plants, supporting level 2 probabilistic risk assessment (PRA) studies. Originally developed as a fast running tool with simplified models for performing probabilistic safety analyses and sensitivity studies, MELCOR now employs largely best estimate models for severe accident phenomena while retaining rich capabilities for performing sensitivity and uncertainty analysis. Extensive user access to fundamental coefficients and parameters of MELCOR underlying physics models combined with the advent of fast and increasingly economical computing power has enabled complex system models like MELCOR to be applied to uncertainty analyses such as those performed using Monte Carlo and Latin Hypercube Sampling (LHS) methods. This paper illustrates two such recent applications of MELCOR 1.8.5 to the uncertainty analysis of hydrogen production and of containment aerosol behavior in severe reactor accidents. The examples illustrated in this paper show both the application of LHS to accommodate computationally intensive accident models and a Monte Carlo analysis involving larger sample size for less computationally intensive analyses to attain better sampling statistics. (author)
Additional details
Publishing Information
- Publisher
- Organisation for Economic Co-Operation and Development - Nuclear Energy Agency - Committee on the Safety of Nuclear Installations
- Imprint Place
- Paris (France)
- Imprint Title
- Evaluation of uncertainties in relation to severe accidents and level-2 probabilistic safety analysis
- Imprint Pagination
- 1002 p.
- Journal Page Range
- p. 398-449
Conference
- Title
- Workshop proceedings evaluation of uncertainties in relation to severe accidents and level-2 probabilistic safety analysis
- Dates
- 7-9 Nov 2005
- Place
- Aix-en-Provence (France)
INIS
- Country of Publication
- France
- Country of Input or Organization
- Nuclear Energy Agency of the OECD (NEA)
- INIS RN
- 39033929
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S37: INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL CHEMISTRY; S42: ENGINEERING;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CHEMICAL REACTION KINETICS; CHEMICAL REACTION YIELD; COMPUTERIZED SIMULATION; CONTAINMENT BUILDINGS; DATA COVARIANCES; DECONTAMINATION; FALLOUT DEPOSITS; FISSION PRODUCT RELEASE; HYDROGEN; LOSS OF COOLANT; M CODES; MELTDOWN; MONTE CARLO METHOD; OXIDATION; PROBABILISTIC ESTIMATION; PWR TYPE REACTORS; RADIOACTIVE AEROSOLS; REACTOR SAFETY; REGRESSION ANALYSIS; SENSITIVITY ANALYSIS; SOURCE TERMS; THERMAL HYDRAULICS; ZIRCALOY
- Descriptors DEC
- ACCIDENTS; AEROSOLS; ALLOYS; BUILDINGS; CALCULATION METHODS; CHEMICAL REACTIONS; CLEANING; COLLOIDS; COMPUTER CODES; CONTAINMENT; DISPERSIONS; ELEMENTS; ENRICHED URANIUM REACTORS; FALLOUT; FLUID MECHANICS; HYDRAULICS; KINETICS; MATHEMATICS; MECHANICS; NONMETALS; POWER REACTORS; REACTION KINETICS; REACTOR ACCIDENTS; REACTORS; SAFETY; SIMULATION; SOLS; STATISTICS; THERMAL REACTORS; TRANSITION ELEMENT ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS; YIELDS; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS
Optional Information
- Notes
- 13 refs.