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Title: BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

Abstract

The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using themore » SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)« less

Authors:
;  [1];  [2];  [1]
  1. Karlsruhe Inst. of Technology (KIT), Inst. for Neutron Physics and Reactor Technology INR, Hermann-vom-Helmholtz-Platz-1, D-76344 Eggenstein-Leopoldshafen (Germany)
  2. Westinghouse Electric Germany GmbH, Mannheim (Germany)
Publication Date:
Research Org.:
American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
OSTI Identifier:
22107754
Resource Type:
Conference
Resource Relation:
Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 12 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; COMPUTER CODES; DATA COVARIANCES; EVALUATION; MONTE CARLO METHOD; NUCLEAR POWER PLANTS; REACTOR KINETICS; RISK ASSESSMENT; SENSITIVITY; THERMAL HYDRAULICS; THREE-DIMENSIONAL CALCULATIONS; TRANSIENTS

Citation Formats

Hartmann, C., Sanchez, V., Tietsch, W., and Stieglitz, R. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification. United States: N. p., 2012. Web.
Hartmann, C., Sanchez, V., Tietsch, W., & Stieglitz, R. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification. United States.
Hartmann, C., Sanchez, V., Tietsch, W., and Stieglitz, R. Sun . "BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification". United States.
@article{osti_22107754,
title = {BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification},
author = {Hartmann, C. and Sanchez, V. and Tietsch, W. and Stieglitz, R.},
abstractNote = {The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2012},
month = {7}
}

Conference:
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