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Title: DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems

Abstract

ANCKVIPRE and the VERA-CS code systems have been successfully applied to predicting PWR core response with respect to departure from nucleate boiling (DNB) at the limiting time step of a postulated hot zero power (HZP) main steam-line break (MSLB) event with and without offsite power available. ANCKVIPRE links the ANC code with the VIPRE-W code. ANC is Westinghouse's multi-dimensional nodal diffusion code for all nuclear core design calculations. It predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. VIPRE-W is the Westinghouse version of a thermalhydraulic subchannel code developed for light water reactor core design applications. The reactor core simulation was based on input of the core boundary conditions from RETRAN system transient and STAR-CCM+ CFD core inlet distribution calculations. To account for the asymmetric power distribution in the reactor core due to the broken steam pipe in one loop and the stuck RCCA, both ANCKVIPRE and the VERA-CS models were set up for the whole reactor core. The advanced modeling and simulations confirmed that the HZP MSLB high-flow case with offsite power available is more limiting with respect to the DNB acceptance criterion, consistent with the limiting case analyzed in the 4-loop plantmore » Safety Analysis Report (SAR). The results also demonstrated that the currently NRC-approved Westinghouse coupled code system such as ANCKVIPRE remains conservative.« less

Authors:
; ; ; ;  [1];  [2]
  1. Westinghouse Electric Company, Cranberry Township, PA 16066 (United States)
  2. Sandia National Laboratories, Albuquerque, NM 87185 (United States)
Publication Date:
Research Org.:
American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
OSTI Identifier:
22765224
Resource Type:
Conference
Resource Relation:
Conference: TOP FUEL 2016: LWR fuels fuels with enhanced safety and performance, Boise, ID (United States), 11-15 Sep 2016; Other Information: Country of input: France; 10 refs.; Related Information: In: TOP FUEL 2016 Proceedings| 1670 p.
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 42 ENGINEERING; A CODES; ASYMMETRY; BOUNDARY CONDITIONS; C CODES; COMPUTERIZED SIMULATION; DEPARTURE NUCLEATE BOILING; EVALUATION; FLUID MECHANICS; NUCLEAR CORES; NUCLEAR INDUSTRY; PIPES; POWER DISTRIBUTION; PWR TYPE REACTORS; REACTOR CORES; SAFETY ANALYSIS; STEAM; STEAM LINE BREAK ACCIDENTS; STEAM LINES; V CODES

Citation Formats

Kucukboyaci, V. N., Sung, Y., Hilton, P. A., Sugimoto, M. A., Huria, H., and Gordon, N. C. DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems. United States: N. p., 2016. Web.
Kucukboyaci, V. N., Sung, Y., Hilton, P. A., Sugimoto, M. A., Huria, H., & Gordon, N. C. DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems. United States.
Kucukboyaci, V. N., Sung, Y., Hilton, P. A., Sugimoto, M. A., Huria, H., and Gordon, N. C. Fri . "DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems". United States.
@article{osti_22765224,
title = {DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems},
author = {Kucukboyaci, V. N. and Sung, Y. and Hilton, P. A. and Sugimoto, M. A. and Huria, H. and Gordon, N. C.},
abstractNote = {ANCKVIPRE and the VERA-CS code systems have been successfully applied to predicting PWR core response with respect to departure from nucleate boiling (DNB) at the limiting time step of a postulated hot zero power (HZP) main steam-line break (MSLB) event with and without offsite power available. ANCKVIPRE links the ANC code with the VIPRE-W code. ANC is Westinghouse's multi-dimensional nodal diffusion code for all nuclear core design calculations. It predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. VIPRE-W is the Westinghouse version of a thermalhydraulic subchannel code developed for light water reactor core design applications. The reactor core simulation was based on input of the core boundary conditions from RETRAN system transient and STAR-CCM+ CFD core inlet distribution calculations. To account for the asymmetric power distribution in the reactor core due to the broken steam pipe in one loop and the stuck RCCA, both ANCKVIPRE and the VERA-CS models were set up for the whole reactor core. The advanced modeling and simulations confirmed that the HZP MSLB high-flow case with offsite power available is more limiting with respect to the DNB acceptance criterion, consistent with the limiting case analyzed in the 4-loop plant Safety Analysis Report (SAR). The results also demonstrated that the currently NRC-approved Westinghouse coupled code system such as ANCKVIPRE remains conservative.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2016},
month = {7}
}

Conference:
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