Sensitivity and Uncertainty Analysis of ARGO-3 Code on the ULOF Event of the 4S Reactor
- Toshiba Corporation, 8, Shinsugita-Cho, Isogo-Ward, Yokohama 235-8523 (Japan)
- Central Research Institute of Electric Power Industry (CRIEPI), 8, Shinsugita-Cho, Isogo-Ward, Yokohama 235-8523 (Japan)
A conceptual design of a sodium cooled fast reactor as a small-decentralized power supply has been performed by CRIEPI and Toshiba Corporation. The reactor named 4S (Super Safe, Small and Simple) has a reflector-controlled U-Zr metallic fuel core. The characteristics of the 4S reactor are non-refueling (core life time: 30 years), a negative coolant temperature reactivity coefficient over 30 years, an integrated and simple reactor structure, and natural circulation decay heat removal systems. An extensive safety analysis for 4S reactor has been conducted for Design Basis Event (DBE) and Beyond Design Basis Event (BDBE) with the transient dynamics code ARGO-3 that has been developed by Toshiba. The purpose of this paper is to show the analytical result that includes the result of the sensitivity and uncertainty analysis for safety of 4S reactor. In this work, first, Anticipated Transient Without Scram (ATWS) is assumed in which both the redundant primary pumps instantaneously stop, which means that a capability of the safety grade coast-down is lost, and the redundant reactor shutdown system fails to scram. Then the coolant temperature in the core outlet and Cumulative Damage Fraction (CDF) are tentatively specified as performance factor for safety in this work. Second, pressure loss coefficient, reactivity feedback and so on are selected as key parameter that affects the analysis result. Third, sensitivity analyses for those parameters are performed. Finally, uncertainty analyses using response surface correlations based on the result of sensitivity analyses are performed. The result on Unprotected Loss of Flow (ULOF), which is one sequence of ATWS, shows that this analysis allows the uncertainty of ARGO-3 to be quantified and the result for the coolant temperature in the core outlet and CDF with ARGO-3 is within the tentative numerical measures. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21021064
- Resource Relation:
- Conference: 2006 International congress on advances in nuclear power plants - ICAPP'06, Reno - Nevada (United States), 4-8 Jun 2006; Other Information: Country of input: France; 8 refs; Related Information: In: Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06, 2734 pages.
- Country of Publication:
- United States
- Language:
- English
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