Depressurization as an accident management strategy to minimize the consequences of direct containment heating

Technical Report · Mon Oct 01 04:00:00 EDT 1990

DOI:https://doi.org/10.2172/6387470· OSTI ID:6387470

Hanson, D J; Golden, D W; Chambers, R; Miller, J D; Hallbert, B P; Dobbe, C A ^[1]

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1. EG and G Idaho, Inc., Idaho Falls, ID (USA)

Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (USA). Div. of Engineering; EG and G Idaho, Inc., Idaho Falls, ID (USA)
• Sponsoring Organization: NRC
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 6387470
• Report Number(s): NUREG/CR-5447; EGG--2574; ON: TI91002467
• Country of Publication: United States
• Language: English

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ACCIDENTS
BOILERS
COMPUTER CALCULATIONS
COMPUTER CODES
CONTAINERS
CONTAINMENT
CONTAINMENT SYSTEMS
CONTROL EQUIPMENT
COOLANTS
COOLING SYSTEMS
DEPRESSURIZATION
ENERGY SYSTEMS
ENGINEERED SAFETY SYSTEMS
EQUIPMENT
FAILURES
FLOW REGULATORS
HIGH TEMPERATURE
PRESSURE VESSELS
PROBABILISTIC ESTIMATION
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR SAFETY
REACTORS
RELIEF VALVES
RISK ASSESSMENT
S CODES
SAFETY
STEAM GENERATORS
VALVES
VAPOR GENERATORS
VENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS