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

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; Sung, Yixing

doi:10.1016/j.nucengdes.2016.09.006

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

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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:

Brown, Cameron S. ^[1]; Zhang, Hongbin ^[2]; Kucukboyaci, Vefa ^[3]; Sung, Yixing ^[3]

North Carolina State Univ., Raleigh, NC (United States)
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Westinghouse Electric Company, Cranberry Township, PA (United States)

Publication Date:: Wed Sep 07 00:00:00 EDT 2016

Research Org.:: Idaho National Lab. (INL), Idaho Falls, ID (United States)

Sponsoring Org.:: USDOE Office of Nuclear Energy (NE)

OSTI Identifier:: 1357754

Alternate Identifier(s):: OSTI ID: 1397899

Report Number(s):: INL/JOU-16-38438
Journal ID: ISSN 0029-5493; PII: S0029549316303247

Grant/Contract Number:: AC07-05ID14517

Resource Type:: 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.

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                    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.  https://doi.org/10.1016/j.nucengdes.2016.09.006

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                    Brown, Cameron S., Zhang, Hongbin, Kucukboyaci, Vefa, and Sung, Yixing. Wed .  
"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.  https://doi.org/10.1016/j.nucengdes.2016.09.006.  https://www.osti.gov/servlets/purl/1357754.

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@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         = {Wed Sep 07 00:00:00 EDT 2016},

  month        = {Wed Sep 07 00:00:00 EDT 2016}

}

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