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Title: MC21/COBRA-IE and VERA-CS multiphysics solutions to VERA core physics benchmark problem #6

Abstract

The Virtual Environment for Reactor Applications (VERA) core physics benchmark problem #6, 3D Hot Full Power (HFP) assembly, from the Consortium for Advanced Simulation of Light Water Reactors (CASL) was simulated using the MC21 continuous energy Monte Carlo code coupled with the COBRA-IE subchannel thermal-hydraulics code using the R5EXEC coupling framework. The converged MC21/COBRA-IE solution was compared to results from CASL's VERA-CS code system, MPACT coupled to COBRA-TF (CTF). MPACT is a three-dimensional (3D) whole core transport code, executed in a 2D/1D approach employing planar method of characteristics (MOC) solutions with SP 3 in the axial direction, and CTF is a subchannel thermal-hydraulics code designed for Light Water Reactor analysis. Eigenvalues agreed within 63 pcm, axially-integrated normalized radial fission distributions agreed within ±0.2% (root mean square (RMS) difference of 0.1%), local volume-averaged fuel pin temperatures agreed within +8.8/-4.3 C (RMS difference of 3.9 C), and local subchannel coolant temperatures agreed within +0.8/-1.5 C (RMS difference of 0.5 C). A sensitivity study to guide tube heat transfer indicated that a statistically-significant increase in reactivity and shift in radial pin power distribution occurred within the assembly when guide tube heating was enabled.

Authors:
; ; ; ORCiD logo; ; ; ;
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility (OLCF)
Sponsoring Org.:
USDOE Office of Science (SC)
OSTI Identifier:
1364065
Alternate Identifier(s):
OSTI ID: 1565664
Grant/Contract Number:  
[AC05-00OR22725]
Resource Type:
Published Article
Journal Name:
Progress in Nuclear Energy
Additional Journal Information:
[Journal Name: Progress in Nuclear Energy Journal Volume: 101 Journal Issue: PC]; Journal ID: ISSN 0149-1970
Publisher:
Elsevier
Country of Publication:
United Kingdom
Language:
English
Subject:
73 NUCLEAR PHYSICS AND RADIATION PHYSICS; Nuclear Science & Technology; Multiphysics; MC21; COBRA-IE; VERA; MPACT; CTF

Citation Formats

Aviles, Brian N., Kelly, Daniel J., Aumiller, David L., Gill, Daniel F., Siebert, Brett W., Godfrey, Andrew T., Collins, Benjamin S., and Salko, Robert K. MC21/COBRA-IE and VERA-CS multiphysics solutions to VERA core physics benchmark problem #6. United Kingdom: N. p., 2017. Web. doi:10.1016/j.pnucene.2017.05.017.
Aviles, Brian N., Kelly, Daniel J., Aumiller, David L., Gill, Daniel F., Siebert, Brett W., Godfrey, Andrew T., Collins, Benjamin S., & Salko, Robert K. MC21/COBRA-IE and VERA-CS multiphysics solutions to VERA core physics benchmark problem #6. United Kingdom. doi:10.1016/j.pnucene.2017.05.017.
Aviles, Brian N., Kelly, Daniel J., Aumiller, David L., Gill, Daniel F., Siebert, Brett W., Godfrey, Andrew T., Collins, Benjamin S., and Salko, Robert K. Wed . "MC21/COBRA-IE and VERA-CS multiphysics solutions to VERA core physics benchmark problem #6". United Kingdom. doi:10.1016/j.pnucene.2017.05.017.
@article{osti_1364065,
title = {MC21/COBRA-IE and VERA-CS multiphysics solutions to VERA core physics benchmark problem #6},
author = {Aviles, Brian N. and Kelly, Daniel J. and Aumiller, David L. and Gill, Daniel F. and Siebert, Brett W. and Godfrey, Andrew T. and Collins, Benjamin S. and Salko, Robert K.},
abstractNote = {The Virtual Environment for Reactor Applications (VERA) core physics benchmark problem #6, 3D Hot Full Power (HFP) assembly, from the Consortium for Advanced Simulation of Light Water Reactors (CASL) was simulated using the MC21 continuous energy Monte Carlo code coupled with the COBRA-IE subchannel thermal-hydraulics code using the R5EXEC coupling framework. The converged MC21/COBRA-IE solution was compared to results from CASL's VERA-CS code system, MPACT coupled to COBRA-TF (CTF). MPACT is a three-dimensional (3D) whole core transport code, executed in a 2D/1D approach employing planar method of characteristics (MOC) solutions with SP3 in the axial direction, and CTF is a subchannel thermal-hydraulics code designed for Light Water Reactor analysis. Eigenvalues agreed within 63 pcm, axially-integrated normalized radial fission distributions agreed within ±0.2% (root mean square (RMS) difference of 0.1%), local volume-averaged fuel pin temperatures agreed within +8.8/-4.3 C (RMS difference of 3.9 C), and local subchannel coolant temperatures agreed within +0.8/-1.5 C (RMS difference of 0.5 C). A sensitivity study to guide tube heat transfer indicated that a statistically-significant increase in reactivity and shift in radial pin power distribution occurred within the assembly when guide tube heating was enabled.},
doi = {10.1016/j.pnucene.2017.05.017},
journal = {Progress in Nuclear Energy},
number = [PC],
volume = [101],
place = {United Kingdom},
year = {2017},
month = {11}
}

Journal Article:
Free Publicly Available Full Text
Publisher's Version of Record
DOI: 10.1016/j.pnucene.2017.05.017

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Cited by: 1 work
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