A multidisciplinary assessment of operation action time for mitigating a postulated accident

Conference · Fri Jan 01 00:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5804918

Buczek, J A ^[1]; Fields, C C; Hightower, III, N T; Morgan, C E ^[2]; Morgan, T B ^[3]; Schneidman, B ^[4]; Swanson, P J ^[5]

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1. ABBGSI, Aiken, SC (United States)
2. Westinghouse Savannah River Company, Aikens, SC (United States)
3. Senergy, Inc., Chattanooga, TN (United States)
4. G. Barri, Inc., Aiken, SC (United States)
5. Concord Associates, Knoxville, TN (United States)

During normal operation of the Savannah River site (SRS) K reactor, heat is transferred from the reactor to the cooling water system (CWS). The river water system (RWS) normally maintains the cooling water inventory in the cooling water (CW) basin. Water is pumped from the CW basin through process water system heat exchangers. The heated water flows into headers that direct water to the effluent sump. Normally, this effluent flows over a weir in the sump, then to the K-area out-fall, and eventually to the Savannah River. An interruption in the RWS supply to tile CW basin is an abnormal condition that has the potential to adversely affect reactor safety. Following this loss of river water to the CW basin, the reactor must be shut down and water conservation established to ensure a 72-h supply of cooling water. This paper briefly describes a methodology developed by the accident management group at the Savannah River site (SRS) for review of the safety analysis and operating/emergency procedures for selected design-basis accidents. The major focus of the review methodology is to identify any discrepancies between the safety analysis and the guidance given to reactor operators that could challenge the safety envelope described in the safety analysis report (SAR). A secondary focus of the methodology is to determine the likelihood that the operator is capable of performing the critical tasks as described in the SAR. The methodology addresses the assumptions made in performing the safety analysis and the results of the analysis of plant response to those conditions. This approach will be used in the review of the existing event-based procedures for selected design-basis accidents at SRS. The review will address the major components, support systems, and essential instruments.

• OSTI ID: 5804918
• Report Number(s): CONF-930601-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 68; Conference: American Nuclear Society (ANS) annual meeting, San Diego, CA (United States), 20-24 Jun 1993; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
K REACTOR
SAFETY ANALYSIS
REACTOR ACCIDENTS
MITIGATION
ECCS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
ENGINEERED SAFETY SYSTEMS
HEAVY WATER MODERATED REACTORS
PRODUCTION REACTORS
REACTOR COMPONENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
SPECIAL PRODUCTION REACTORS
210400* - Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
220900 - Nuclear Reactor Technology- Reactor Safety