A computer model for the transient analysis of compact research reactors with plate type fuel

Conference · Tue Mar 01 00:00:00 EST 1994

OSTI ID:10131029

Sofu, T ^[1]; Dodds, H L ^[2]

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1. Argonne National Lab., IL (United States)
2. Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering

A coupled neutronics and core thermal-hydraulic performance model has been developed for the analysis of plate type U-Al fueled high-flux research reactor transients. The model includes point neutron kinetics, one-dimensional, non-homogeneous, equilibrium two-phase flow and beat transfer with provision for subcooled boiling, and spatially averaged one-dimensional beat conduction. The feedback from core regions other than the fuel elements is included by employing a lumped parameter approach. Partial differential equations are discretized in space and the combined equation set representing the model is converted to an initial value problem. A variable-order, variable-time-step time advancement scheme is used to solve these ordinary differential equations. The model is verified through comparisons with two other computer code results and partially validated against SPERT-II tests. It is also used to analyze a series of HFIR reactivity transients.

• Research Organization: Argonne National Lab., IL (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 10131029
• Report Number(s): ANL/RA/CP-79957; CONF-940402-12; ON: DE94007675
• Resource Relation: Conference: International topical meeting on the safety of advanced reactors,Pittsburgh, PA (United States),18-20 Apr 1994; Other Information: PBD: [1994]
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
RESEARCH REACTORS
FUEL PLATES
TRANSIENTS
COMPUTER CODES
HFIR REACTOR
ATR REACTOR
HFBR REACTOR
MATHEMATICAL MODELS
HEAT TRANSFER
HYDRAULICS
REACTOR PHYSICS
REACTOR SAFETY
220900
220600
220100
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
THEORY AND CALCULATION