RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident
A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.
- Research Organization:
- Brookhaven National Lab., Upton, NY (USA)
- Sponsoring Organization:
- DOE/ER
- DOE Contract Number:
- AC02-76CH00016
- OSTI ID:
- 7027691
- Report Number(s):
- BNL-52243; ON: DE90014716
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
AFTER-HEAT REMOVAL
BLACKOUTS
COMPUTER CODES
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
HFBR REACTOR
HYDRAULICS
LOSS OF COOLANT
MECHANICS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
TANK TYPE REACTORS
THERMAL REACTORS
TRANSIENTS