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Title: Development of a detailed core flow analysis code for prismatic fuel reactors

Abstract

The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR.

Authors:
Publication Date:
OSTI Identifier:
5603184
Alternate Identifier(s):
OSTI ID: 5603184
Report Number(s):
CONF-901101--
Journal ID: ISSN 0003-018X; CODEN: TANSA
Resource Type:
Conference
Resource Relation:
Journal Name: Transactions of the American Nuclear Society; (USA); Journal Volume: 62; Conference: American Nuclear Society (ANS) winter meeting, Washington, DC (USA), 11-15 Nov 1990
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; HTGR TYPE REACTORS; REACTOR CORES; REACTOR SAFETY; B CODES; GAS FLOW; HEAT TRANSFER; PRISMATIC CONFIGURATION; ACCURACY; COMPUTER CODES; COMPUTERIZED SIMULATION; CONTROL ELEMENTS; CONVERGENCE; FLOW BLOCKAGE; FUEL ELEMENTS; HELIUM; ITERATIVE METHODS; JACOBIAN FUNCTION; MODULAR STRUCTURES; NEUTRON ABSORBERS; NONLINEAR PROBLEMS; PERFORMANCE; STEADY-STATE CONDITIONS; VALIDATION; VERIFICATION; CONFIGURATION; ELEMENTS; ENERGY TRANSFER; FLUID FLOW; FLUIDS; FUNCTIONS; GAS COOLED REACTORS; GASES; GRAPHITE MODERATED REACTORS; NONMETALS; RARE GASES; REACTOR COMPONENTS; REACTORS; SAFETY; SIMULATION; TESTING 220900* -- Nuclear Reactor Technology-- Reactor Safety; 210300 -- Power Reactors, Nonbreeding, Graphite Moderated

Citation Formats

Bennett, R.G. Development of a detailed core flow analysis code for prismatic fuel reactors. United States: N. p., 1990. Web.
Bennett, R.G. Development of a detailed core flow analysis code for prismatic fuel reactors. United States.
Bennett, R.G. Mon . "Development of a detailed core flow analysis code for prismatic fuel reactors". United States. doi:.
@article{osti_5603184,
title = {Development of a detailed core flow analysis code for prismatic fuel reactors},
author = {Bennett, R.G.},
abstractNote = {The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR.},
doi = {},
journal = {Transactions of the American Nuclear Society; (USA)},
number = ,
volume = 62,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 1990},
month = {Mon Jan 01 00:00:00 EST 1990}
}

Conference:
Other availability
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