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Title: VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

Abstract

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, themore » results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Authors:
 [1];  [2];  [2];  [2];  [2]
  1. ORNL
  2. Westinghouse Electric Company, Cranberry Township
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Consortium for Advanced Simulation of LWRs (CASL)
Sponsoring Org.:
USDOE
OSTI Identifier:
1266852
DOE Contract Number:  
AC05-00OR22725
Resource Type:
Conference
Resource Relation:
Conference: The International Conference on Nuclear Engineering (ICONE) 24th, Charlotte,, NC, USA, 20160626, 20160630
Country of Publication:
United States
Language:
English
Subject:
CTF; DNB

Citation Formats

Salko, Robert K, Sung, Yixing, Kucukboyaci, Vefa, Xu, Yiban, and Cao, Liping. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB. United States: N. p., 2016. Web.
Salko, Robert K, Sung, Yixing, Kucukboyaci, Vefa, Xu, Yiban, & Cao, Liping. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB. United States.
Salko, Robert K, Sung, Yixing, Kucukboyaci, Vefa, Xu, Yiban, and Cao, Liping. Fri . "VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB". United States.
@article{osti_1266852,
title = {VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB},
author = {Salko, Robert K and Sung, Yixing and Kucukboyaci, Vefa and Xu, Yiban and Cao, Liping},
abstractNote = {The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2016},
month = {1}
}

Conference:
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