SIMMER-IV application to safety assessment of severe accident in a small SFR
2
조회
0
다운로드
0
인용
0
참고문헌
키워드
초록
A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.
Abstract
A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.
참고문헌 (12)
- Defense in Depth in Nuclear Safety, "International Nuclear Safety Advisory Group", IAEA, ~, pp.INSAG-10 (1996)
- S. Kondo, "SIMMER-III: an Advanced Computer Program for LMFBR Severe Accident Analysis", ANP’92 International Conference on Design and Safety of Advanced Nuclear Power Plants (1992)
- SIMMER-IV: A Three Dimensional Computer Program for LMFR Core Disruptive Accident Analysis, "H. Yamano", Japan Nuclear Cycle Development Institute, ~, pp.JNC TN9400 (2003)
- Y. Tobita, "The development of SIMMER-III, an advanced computer program for LMFR safety analysis, and its application to sodium experiments", Nucl. Tech, 153, pp.245~255 (2006)
- Phase 1 Code Assessment of SIMMER-III; A Computer Program for LMFR Core Disruptive Accident Analysis, "Sa Kondo", JAEA-Research, 009~ (2019)
- Phase 2 Code Assessment of SIMMER-III; A Computer Program for LMFR Core Disruptive Accident Analysis, "Sa Kondo", JNC, 105~, pp.JNC TN9400 (2000)
- Sa Kondo, "IMMER-III Code and Space-Energy Dependent Dynamics of Fast Reactor", 7th Topical Meeting on Nuclear Code Development (1991)
- THREEDANT: A Code to Perform Three-Dimensional, Neutral Particle Transport Calculations, "R. E. Alcouffe", Los Alamos National Lab, ~, pp.No. LA-UR–94-3259 (1994)
- T. Nakagawa, "Japanese evaluated nuclear data library version 3 revision-2: JENDL-3.2", J. Nucl. Sci. Technol. (Tokyo, Jpn.), 32, pp.1259~1271 (1995)
- P. Anzieu, "The CORECT-II Fuel-Coolant Interaction Experiments: Interpretation and Subassembly Accident Model", Proc. LMFBR Safety Topical Mtg (1982)
- B. W. Spencer, "Results of Recent Reactor-Materials Tests on Dispersal of Oxide Fuel from a Disrupted Core", Proc. Int. Topl. Mtg. Fast Reactor Safety (1985)
- K. Konishi, "The EAGLE project to eliminate the recriticality issue of fast reactors - progress and results of in-pile tests", Proc. NTHAS5, pp.465~471 (2006)