Analysis of N Reactor anticipated transients without scram

Technical Report · Sat Jul 01 04:00:00 EDT 1989

OSTI ID:7228690

Sathyanarayana, K ^[1]; Ogden, D M ^[2]

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1. Pacific Northwest Lab., Richland, WA (USA)
2. Westinghouse Hanford Co., Richland, WA (USA)

This report documents the N Reactor thermal-hydraulic system response for the anticipated transients without SCRAM (ATWS). The N Reactor is a graphite-moderated, water-cooled pressure-tube reactor. The primary coolant is pumped through 1003 horizontal pressure-tubes which contain two concentric tubular metallic fuel elements. The calculations were performed using RELAP5/MOD2 computer code. RELAP5/MOD2 is a best-estimate, two-phase, full six-equation, nonequilibrium, and nonhomogeneous thermal-hydraulic computer code designed for pressurized water reactor transients. A point-reactor kinetics model with reactivity feedback was employed to compute the power generated in the reactor core. The transients simulated were 200% double-ended guillotine break in the cold-leg manifold and the station blackout with the simultaneous failure of SCRAM and emergency core cooling system (ECCS) and the loss of ac power. In all the transients, the loss of high pressure injection system (HPI), the complete loss of graphite and shield cooling system (GSCS) and the isolation of steam generator secondary were included. Also, the ATWS transient, due to the 200% guillotine break of the cold-leg manifold assuming the availability of ECCS from three high-lift diesel pumps operating at full capacity and the loss of ac power, was analyzed. These three transients were first evaluated using the void feedback reactivity having an absolute level of {minus}167 mk corresponding to an infinite lattice MK IV fuel. Then these transients were reevaluated using a more realistic void feedback reactivity having an absolute level of {minus}70 mk. The report includes brief descriptions of the N Reactor, the computer code, the thermal-hydraulic plant model used in the simulations and the discussion of results. 12 refs., 113 figs., 13 tabs.

• Research Organization: Westinghouse Hanford Co., Richland, WA (USA)
• Sponsoring Organization: DOE/DP
• DOE Contract Number: AC06-87RL10930
• OSTI ID: 7228690
• Report Number(s): WHC-SP-0457; ON: DE90009485
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220700 -- Nuclear Reactor Technology-- Plutonium & Isotope Production Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
ATWS
BLACKOUTS
BOILERS
COMPUTER CODES
CONTROL EQUIPMENT
COOLING SYSTEMS
ECCS
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
EQUIPMENT
FLOW REGULATORS
FLUID MECHANICS
GRAPHITE MODERATED REACTORS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
LWGR TYPE REACTORS
MECHANICS
N-REACTOR
PLUTONIUM PRODUCTION REACTORS
POWER REACTORS
PRESSURE TUBES
PRESSURIZERS
PRODUCTION REACTORS
PUMPS
R CODES
REACTIVITY
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
STEAM GENERATORS
TRANSIENTS
TUBES
VALVES
VAPOR GENERATORS
VOID FRACTION
WATER COOLED REACTORS