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Title: Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

Abstract

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by a system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.

Authors:
 [1];  [2];  [3];  [3]
  1. North Carolina State Univ., Raleigh, NC (United States)
  2. Idaho National Lab. (INL), Idaho Falls, ID (United States)
  3. Westinghouse Electric Company, Cranberry Township, PA (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1357754
Report Number(s):
INL/JOU-16-38438
Journal ID: ISSN 0029-5493; PII: S0029549316303247
Grant/Contract Number:
AC07-05ID14517
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Nuclear Engineering and Design
Additional Journal Information:
Journal Volume: 309; Journal ID: ISSN 0029-5493
Publisher:
Elsevier
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; VERA-CS, BEPU, MSLB

Citation Formats

Brown, Cameron S., Zhang, Hongbin, Kucukboyaci, Vefa, and Sung, Yixing. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event. United States: N. p., 2016. Web. doi:10.1016/j.nucengdes.2016.09.006.
Brown, Cameron S., Zhang, Hongbin, Kucukboyaci, Vefa, & Sung, Yixing. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event. United States. doi:10.1016/j.nucengdes.2016.09.006.
Brown, Cameron S., Zhang, Hongbin, Kucukboyaci, Vefa, and Sung, Yixing. 2016. "Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event". United States. doi:10.1016/j.nucengdes.2016.09.006. https://www.osti.gov/servlets/purl/1357754.
@article{osti_1357754,
title = {Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event},
author = {Brown, Cameron S. and Zhang, Hongbin and Kucukboyaci, Vefa and Sung, Yixing},
abstractNote = {VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by a system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.},
doi = {10.1016/j.nucengdes.2016.09.006},
journal = {Nuclear Engineering and Design},
number = ,
volume = 309,
place = {United States},
year = 2016,
month = 9
}

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  • Virginia Power operates four nuclear reactors, two units each at the Surry and North Anna Power stations. The original operating licenses were based on acceptable analysis results of the accidents in the final safety analysis report (FSAR). The assumptions of these analyses must be verified on a reload basis. Included in these FSAR accidents is the main steam-line-break (MSLB) event. The plant FSARs describe the MSLB analyses, which is summarized as follows. The plant is assumed to be at hot zero power at end of life, when the moderator temperature coefficient (MTC) is most negative. The MSLB rapidly cools themore » secondary side, followed by a primary cooldown in one loop. The surge of cold water into the core, coupled with the negative MTC, results in high local power factors, which in turn can result in a violation of the departure from nucleate boiling ratio (DNBR) limit. The three-dimensional power distribution is calculated at several key state points. These distributions are then subjected to core thermal-hydraulic analysis by the COBRA code. The W-3 correlation is used to calculate the state-point DNBRs. Both the physics and the DNBR calculations have been repeated on a reload basis. As a result, Virginia Power has accumulated a reasonably large data base of MSLB DNBRs for both Surry and North Anna. Virginia Power now uses the power peaking factors criterion to verify that the MSLB analysis remains bounding on a reload basis.« less
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