Simulation of the loss of RHR during midloop operations and the role of steam generators in decay heat removal

Conference · Wed Jan 01 00:00:00 EST 1992 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6982098

Raja, L L; Banerjee, S; Hassan, Y A ^[1]

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1. Texas A M Univ., College Station (United States)

Loss of residual heat removal (RHR) during midloop operations was simulated using the RELAP5/MOD3 thermal-hydraulic code for a typical four-loop pressurized water reactor (PWR) under reduced inventory level. Two cases are considered here: one for an intact reactor coolant system with no vents and the other for an open system with a vent in the pressurizer. The effect of air on the transients was studied, unlike the RETRAN analysis of core boiling during midloop operations performed by Fujita and Rice, which did not analyze the presence of air in the system. The steam generators have water in the secondary covering the U-tubes. The system gets pressurized once water starts boiling in the core with higher system pressures for the vent-closed case. Reflux condensation occurs in the U-tubes aiding decay heat removal and preventing complete uncovery of the core. Sudden pressurization of the hot leg and vessel upper head causes the reactor vessel to act as a manometer reducing the core level and raising the downcomer level. Fuel centerline and clad temperatures are below safety limits throughout the transients.

• OSTI ID: 6982098
• Report Number(s): CONF-921102-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 66; Conference: Joint American Nuclear Society (ANS)/European Nuclear Society (ENS) international meeting on fifty years of controlled nuclear chain reaction: past, present, and future, Chicago, IL (United States), 15-20 Nov 1992; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PRIMARY COOLANT CIRCUITS
TRANSIENTS
PWR TYPE REACTORS
STEAM GENERATORS
AFTER-HEAT REMOVAL
HYDRAULICS
R CODES
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
RHR SYSTEMS
SIMULATION
THERMAL ANALYSIS
BOILERS
COMPUTER CODES
COOLING SYSTEMS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
FLUID MECHANICS
MECHANICS
POWER REACTORS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
REMOVAL
SAFETY
THERMAL REACTORS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled