Analysis of production reactor response during a postulated Loss-of-River Water event using CONTAIN/SR [electronic resource].

Published: Washington, D.C. : United States. Dept. of Energy, 1992.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
Physical Description: Pages: (19 pages) : digital, PDF file
Additional Creators: Westinghouse Savannah River Company, United States. Department of Energy, and United States. Department of Energy. Office of Scientific and Technical Information

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Summary

This report discusses the CONTAIN/SR computer code, developed at the Savannah River Technology Center and Sandia National Laboratories for Probabilistic Safety Assessment (PSA) applications, which is used to analyze K Reactor plant conditions following a design basis earthquake to assist post-accident recovery planning. The postulated event, a Loss-of-River Water (LORW) accident, requires analysis of the K Reactor confinement system assuming seismic event-caused loss of forced air flow through Radiologically Controlled (RCAs) and other building areas, including adjoining personnel and auxiliary equipment zones. The CONTAIN/SR code calculations predict the expected environment in the K Reactor building with a seismically-qualified flow path for natural circulation, under design basis conditions specifying a 50 gal/min leak of tritiated heavy water. Despite loss of active fan flow, preferential air flow patterns are calculated to flow from clean'' areas towards the RCAs. Ventilation characteristics of the building reduce tritiated water vapor concentrations to habitable levels, assuming plastic suits and clean breathing air supplies are available. Unprotected dose rates to recovery workers in the heat exchanger zone of the building will range from 120 mrem/hour to 780 mrem/hour, depending on evaporation conditions near spilled heavy-water pools. It is concluded habitability issues for recovery are not driven by temperature concerns in reactor building zones. However, the results indicate radiological suits with cool air supplies will assure adequate conditions for operators and recovery teams, and mitigate tritium uptake hazards from splashing or other direct contact mechanisms.

Report Numbers

E 1.99:wsrc-ms-92-026
E 1.99: conf-930352--6
conf-930352--6
wsrc-ms-92-026

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Note

Published through SciTech Connect.
06/01/1992.
"wsrc-ms-92-026"
" conf-930352--6"
"DE93002689"
2. international conference on nuclear engineering, San Francisco, CA (United States), 21-24 Mar 1993.
Jenkins, T.B.; Wooten, L.A.; O'Kula, K.R.

Funding Information

AC09-89SR18035

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