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

Abstract

The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. 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 Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs.

Authors:
Publication Date:
Research Org.:
EG and G Idaho, Inc., Idaho Falls, ID (USA)
Sponsoring Org.:
DOE/NE
OSTI Identifier:
6645795
Alternate Identifier(s):
OSTI ID: 6645795; Legacy ID: DE90013092
Report Number(s):
EGG-M-90249; CONF-901101--7
ON: DE90013092; TRN: 91-010567
DOE Contract Number:
AC07-76ID01570
Resource Type:
Conference
Resource Relation:
Conference: American Nuclear Society winter meeting, Washington, DC (USA), 11-15 Nov 1990
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; HELIUM; FLUID FLOW; HTGR TYPE REACTORS; REACTOR COOLING SYSTEMS; B CODES; COMPUTER CODES; FLOW MODELS; FLUID MECHANICS; FUEL ASSEMBLIES; REACTOR CHANNELS; REACTOR CORES; COOLING SYSTEMS; ELEMENTS; ENERGY SYSTEMS; FLUIDS; GAS COOLED REACTORS; GASES; GRAPHITE MODERATED REACTORS; MATHEMATICAL MODELS; MECHANICS; NONMETALS; RARE GASES; REACTOR COMPONENTS; REACTORS 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:. https://www.osti.gov/servlets/purl/6645795.
@article{osti_6645795,
title = {Development of a detailed core flow analysis code for prismatic fuel reactors},
author = {Bennett, R.G.},
abstractNote = {The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. 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 Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 1990},
month = {Mon Jan 01 00:00:00 EST 1990}
}

Conference:
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  • 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 ofmore » 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.« less
  • No abstract prepared.
  • VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. Itmore » requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)« less
  • Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was analyzed using MCNP, ORIGEN2 and MONTEBURNS codes, considering a homogeneous mixture for the core material. Promising results showed that with standard fuel material of low enrichment uranium, high thermal fluxes are delivered by this core in the outer reflector. Inmore » addition, the life cycle, which is a limiting parameter regarding the compactness of the core, was calculated to be 41 days. In this paper, we report the results of recent developments in the design of this core. A detailed modeling of a suitable fuel element is performed with MCNP. Neutron fluxes are recalculated to assure that the same levels are achieved with this more detailed description. In this regard, three different zones with different flux levels were identified: a high neutron flux zone (4.0 E14 n m{sup -2}s{sup -1}), a moderate thermal neutron flux zone (2.5 E14 n m{sup -2}s{sup -1}), and a low thermal flux zone (1.0 E14 n m{sup -2}s{sup -1}). Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature for safe operating conditions. (authors)« less