Application of (transient reactor analysis code) TRAC to the Savannah River Reactor PRA (probabilistic risk assessment)
- SAIC, Albuquerque, NM (USA)
Calculation of the thermal-hydraulic response of the Savannah River site (SRS) reactors to certain transients has been undertaken in support of the ongoing level-3 probabilistic risk assessment (PRA). Analysis of the accident progression following the onset of fuel damage is part of that assessment (referred to as the level-2 portion). The level-2 assessment begins from an evaluation of the frequencies of postulated core damage accidents and an assessment, for each accident with significant frequency, of the conditions in the primary system and confinement when core damage begins. Core damage frequencies have been obtained from a completed level-1 PRA. Insights into the state of the primary system have been obtained from the melt-initiation studies described in this paper. To carry out these studies, the transient reactor analysis code (TRAC) was employed. Simulation of severe transients, involving multiple failures of engineered safety features, was required. The objective of these studies was threefold: (1) to determine the primary system water level, pressure, and temperature at the incipience of core damage; (2) to estimate the time available for operator intervention to prevent core damage; (3) to gain insight into realistic success criteria.
- OSTI ID:
- 6178342
- Report Number(s):
- CONF-900608-; CODEN: TANSA; TRN: 91-008361
- Journal Information:
- Transactions of the American Nuclear Society; (USA), Vol. 61; Conference: American Nuclear Society (ANS) annual meeting, Nashville, TN (USA), 10-14 Jun 1990; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
REACTOR SAFETY
T CODES
SPECIAL PRODUCTION REACTORS
RISK ASSESSMENT
COMPUTER CODES
COMPUTERIZED SIMULATION
EMERGENCY PLANS
ENGINEERED SAFETY SYSTEMS
HEAT TRANSFER
HYDRAULICS
LANL
PRIMARY COOLANT CIRCUITS
PROBABILITY
REACTOR CORE DISRUPTION
REACTOR OPERATORS
SAVANNAH RIVER PLANT
TRANSIENTS
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
MECHANICS
NATIONAL ORGANIZATIONS
PERSONNEL
PRODUCTION REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
SAFETY
SIMULATION
US AEC
US DOE
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220900* - Nuclear Reactor Technology- Reactor Safety
220700 - Nuclear Reactor Technology- Plutonium & Isotope Production Reactors