System code requirements for safety analysis of SBWR
- GE, Wilmington, DE (United States)
- GE, San Jose, CA (United States)
The simplified boiling water reactor (SBWR) being developed by General Electric Nuclear Energy is an advanced boiling water reactor relying on natural circulation during normal operation and passive safety features. The major elements of the passive safety features are the automatic depressurization of the reactor pressure vessel (RPV) following a loss-of-coolant accident (LOCA) through safety/relief valves and depressurization valves, the gravity-driven coolant system (GDCS), and the passive containment cooling system (PCCS) for residual heat removal. These passive safety systems, although based on existing technology, have generated new requirements for the computer codes used in safety and design analysis. TRACG is the computer code used for safety and design analysis for the SBWR.
- OSTI ID:
- 89168
- Report Number(s):
- CONF-941102-; ISSN 0003-018X; TRN: 95:004215-0242
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 71; Conference: Winter meeting of the American Nuclear Society (ANS), Washington, DC (United States), 13-18 Nov 1994; Other Information: PBD: 1994
- Country of Publication:
- United States
- Language:
- English
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