Uncertainty analysis of minimum vessel liquid inventory during a small-break LOCA in a B W Plant: An application of the CSAU methodology using the RELAP5/MOD3 computer code

Technical Report · Tue Dec 01 00:00:00 EST 1992

DOI:https://doi.org/10.2172/6771672· OSTI ID:6771672

Ortiz, M G; Ghan, L S ^[1]

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

The Nuclear Regulatory Commission (NRC) revised the emergency core cooling system licensing rule to allow the use of best estimate computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability, and Uncertainty (CSAU) to evaluate best estimate code uncertainties. The objective of this work was to adapt and demonstrate the CSAU methodology for a small-break loss-of-coolant accident (SBLOCA) in a Pressurized Water Reactor of Babcock Wilcox Company lowered loop design using RELAP5/MOD3 as the simulation tool. The CSAU methodology was successfully demonstrated for the new set of variants defined in this project (scenario, plant design, code). However, the robustness of the reactor design to this SBLOCA scenario limits the applicability of the specific results to other plants or scenarios. Several aspects of the code were not exercised because the conditions of the transient never reached enough severity. The plant operator proved to be a determining factor in the course of the transient scenario, and steps were taken to include the operator in the model, simulation, and analyses.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research; EG and G Idaho, Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 6771672
• Report Number(s): NUREG/CR-5818; EGG-2665; ON: TI93006225
• Country of Publication: United States
• Language: English

Related Subjects

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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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LOSS OF COOLANT
COMPUTERIZED SIMULATION
PWR TYPE REACTORS
C CODES
DATA COVARIANCES
ECCS
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REACTOR COOLING SYSTEMS
REACTOR LICENSING
REACTOR OPERATORS
ACCIDENTS
ANIMALS
COMPUTER CODES
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ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
LICENSING
MAMMALS
MAN
PERSONNEL
POWER REACTORS
PRIMATES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SIMULATION
THERMAL REACTORS
VERTEBRATES
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
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990200 - Mathematics & Computers