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Title: Computational Fluid Dynamic Analysis of Core Bypass Flow Phenomena in a Prismatic VHTR

Abstract

The core bypass flow in a prismatic very high temperature gas-cooled reactor (VHTR) is one of the important design considerations which impacts considerably on the integrity of reactor core internals including operating fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) will be affected by the bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to understand better the bypass flow phenomenon and establish the evaluation method in the reactor core using commercial CFD code FLUENT. Parametric calculations changing several factors in a on-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the flow and temperature distributions in the prismatic core. The factors inlcude inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-inducedmore » shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block that may have repurcussions on the structural integrity of the block and on the neutronics. These results indicate that bypass flow has a significant effect on hot spots in the core and on the temperature of jets flowing from the core into the lower plenum.« less

Authors:
; ;
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
DOE - NE
OSTI Identifier:
989273
Report Number(s):
INL/JOU-09-16775
Journal ID: ISSN 0306-4549; TRN: US1007059
DOE Contract Number:  
DE-AC07-05ID14517
Resource Type:
Journal Article
Journal Name:
Annals of Nuclear Energy
Additional Journal Information:
Journal Volume: 37; Journal Issue: 9; Journal ID: ISSN 0306-4549
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; COMPUTERIZED SIMULATION; COOLANTS; FLUID MECHANICS; GAS COOLED REACTORS; GAS FLOW; GEOMETRY; GRAPHITE; HOT SPOTS; INTERSTITIALS; IRRADIATION; LIFETIME; MANUFACTURING; REACTOR CORES; SHRINKAGE; TEMPERATURE DISTRIBUTION; TEMPERATURE GRADIENTS; THERMAL EXPANSION; TURBULENCE; CFD,; core bypass flow; prismatic,; VHTR,

Citation Formats

Sato, Hiroyuki, Johnson, Richard W, and Schultz, Richard R. Computational Fluid Dynamic Analysis of Core Bypass Flow Phenomena in a Prismatic VHTR. United States: N. p., 2010. Web. doi:10.1016/j.anucene.2010.04.021.
Sato, Hiroyuki, Johnson, Richard W, & Schultz, Richard R. Computational Fluid Dynamic Analysis of Core Bypass Flow Phenomena in a Prismatic VHTR. United States. https://doi.org/10.1016/j.anucene.2010.04.021
Sato, Hiroyuki, Johnson, Richard W, and Schultz, Richard R. 2010. "Computational Fluid Dynamic Analysis of Core Bypass Flow Phenomena in a Prismatic VHTR". United States. https://doi.org/10.1016/j.anucene.2010.04.021.
@article{osti_989273,
title = {Computational Fluid Dynamic Analysis of Core Bypass Flow Phenomena in a Prismatic VHTR},
author = {Sato, Hiroyuki and Johnson, Richard W and Schultz, Richard R},
abstractNote = {The core bypass flow in a prismatic very high temperature gas-cooled reactor (VHTR) is one of the important design considerations which impacts considerably on the integrity of reactor core internals including operating fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) will be affected by the bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to understand better the bypass flow phenomenon and establish the evaluation method in the reactor core using commercial CFD code FLUENT. Parametric calculations changing several factors in a on-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the flow and temperature distributions in the prismatic core. The factors inlcude inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block that may have repurcussions on the structural integrity of the block and on the neutronics. These results indicate that bypass flow has a significant effect on hot spots in the core and on the temperature of jets flowing from the core into the lower plenum.},
doi = {10.1016/j.anucene.2010.04.021},
url = {https://www.osti.gov/biblio/989273}, journal = {Annals of Nuclear Energy},
issn = {0306-4549},
number = 9,
volume = 37,
place = {United States},
year = {2010},
month = {9}
}