Analysis of high-pressure boiloff situation during an MSIV closure ATWS in a typical BWR/4

Conference · Wed Jan 01 00:00:00 EST 1986 · Trans. Am. Nucl. Soc.; (United States)

OSTI ID:6623650

Neymotin, L Y; Slovik, G C; Saha, P

An anticipated transient without scram (ATWS) is recognized as one of the boiling water reactor (BWR) accident sequences potentially leading to core damage. Of all the various ATWS initiating events, the main steam isolation valve (MSIV) closure ATWS is the most severe, because of its relatively high frequency of occurrence and its challenge to the residual heat removal and containment integrity systems. Although under investigation for quite a long period of time, different aspects of this type of transient are still being analyzed. The final outcome of these studies should be a well-defined set of recommendations for the plant operator to mitigate an ATWS accident. The objective of this paper is to provide a best estimate analysis of the MSIV closure ATWS in the Browns Ferry Unit 1 BWR with Mark-1 containment. The calculations have been performed using the RAMONA-3B code which as a three-dimensional neutron kinetics model coupled with one-dimensional four-equation, nonhomogeneous, nonequilibrium thermal hydraulics. The code also allows for one-dimensional neutronic core representation. The one-dimensional capability of the code has been employed in this calculation since a thorough sensitivity study showed that for a full ATWS, a one-dimensional neutron kinetics adequately describes the core behavior. The calculation described in the paper was started from a steady-state fuel condition corresponding to the end of cycle 5 of the Browns Ferry reactor.

• Research Organization: Brookhaven National Lab., Upton, NY
• OSTI ID: 6623650
• Report Number(s): CONF-860610-; TRN: 87-016350
• Journal Information: Trans. Am. Nucl. Soc.; (United States), Vol. 52; Conference: American Nuclear Society annual meeting, Reno, NV, USA, 15 Jun 1986
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ATWS
HEAT TRANSFER
HYDRAULICS
BROWNS FERRY-1 REACTOR
REACTOR SAFETY
AFTER-HEAT REMOVAL
COMPUTER CODES
COMPUTERIZED SIMULATION
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LOSS OF COOLANT
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REACTOR ACCIDENTS
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TRANSIENTS
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ENERGY TRANSFER
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NATIONAL ORGANIZATIONS
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220900* - Nuclear Reactor Technology- Reactor Safety
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