Application of two different similarity laws for the RVACS design
0
조회
0
다운로드
0
인용
0
참고문헌
키워드
초록
The RVACS is a versatile and robust safety system driven by two natural circulations: in-vessel coolant and ex-vessel air. To observe interaction between the two natural circulations, SINCRO-IT facility was designed with two different similarity laws simultaneously. Bo’ based similarity law was employed for the in-vessel, while Ishii’s similarity law for the ex-vessel excluding the radiation. Compared to the prototype, the sodium and air system, SINCRO-IT was designed with Wood’s metal and air, having 1:4 of the length reduction, and 1.68:1 of the time scale ratio. For the steady state, RV temperature limit was violated at 0.8% of the decay heat, while the sodium boiling was predicted at 1.3%. It showed good accordance with the system code, TRACE. For an arbitrary re-criticality scenario with RVACS solitary operation, sodium boiling was predicted at 25,100 s after power increase from 1.0 to 2.0%, while the system code showed 30,300. Maximum temperature discrepancy between the experiments and system code was 4.2%. The design and methodology were validated by the system code TRACE in terms of the convection, and simultaneously, the system code was validated against the simulating experiments SINCRO-IT. The validated RVACS model could be imported to further accident analysis.
Abstract
The RVACS is a versatile and robust safety system driven by two natural circulations: in-vessel coolant and ex-vessel air. To observe interaction between the two natural circulations, SINCRO-IT facility was designed with two different similarity laws simultaneously. Bo’ based similarity law was employed for the in-vessel, while Ishii’s similarity law for the ex-vessel excluding the radiation. Compared to the prototype, the sodium and air system, SINCRO-IT was designed with Wood’s metal and air, having 1:4 of the length reduction, and 1.68:1 of the time scale ratio. For the steady state, RV temperature limit was violated at 0.8% of the decay heat, while the sodium boiling was predicted at 1.3%. It showed good accordance with the system code, TRACE. For an arbitrary re-criticality scenario with RVACS solitary operation, sodium boiling was predicted at 25,100 s after power increase from 1.0 to 2.0%, while the system code showed 30,300. Maximum temperature discrepancy between the experiments and system code was 4.2%. The design and methodology were validated by the system code TRACE in terms of the convection, and simultaneously, the system code was validated against the simulating experiments SINCRO-IT. The validated RVACS model could be imported to further accident analysis.
참고문헌 (27)
- T. L. Schulz, "Westinghouse AP1000 advanced passive plant", Nuclear Engineering and Design, 236, pp.1547~1557 (2006)
- K. M. Kim, "Analysis of natural circulation behaviors and flow instability of passive containment cooling system design for advanced PWR using MARS-KS code", International Journal of Heat and Mass Transfer, 147, pp.118982~ (2020)
- M. Sawada, "Experiment and analysis of natural convection characteristics in the experimental fast reactor JOYO", Nuclear Engineering and Design, 120, pp.341~347 (1990)
- H. Mochizuki, "Analyses of decay heat removal tests with PRACS and DRACS of a sodium loop using the NETFLOWþþ code", Nuclear Engineering and Design, 322, pp.345~359 (2017)
- T. Mihara, "ASTRID Nuclear Island Design : Advances in French-Japanese Joint Team Development of Decay Heat Removal Systems", ICAPP 2016 - International Congress on Advances in Nuclear Power Plants (2016)
- 유재운, "Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea", Nuclear Engineering and Technology, 48(5), pp.1059~1070 (2016), 한국원자력학회
- A. Alemberti, "Overview of lead cooled fast reactor activities", Progress in Nuclear Energy, 77, pp.300~307 (2014)
- F. Giannetti, "Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors", Nuclear Engineering and Design, 305, pp.168~178 (2016)
- K. V. Tichelen, "Advanced liquid-metal thermal-hydraulic Research for MYRRHA", Nuclear Technology, 206, pp.150~163 (2020)
- S. Grewal, "Water simulation of sodium reactors", Chemical Engineering Communications, 17, pp.343~360 (1982)
- Y. Eguchi, "Quantitative prediction of natural circulation in an LMFR with a similarity law and a water test", Nuclear Engineering and Design, 178, pp.295~307 (1997)
- 이민호, "Experimental validation of simulating natural circulation of liquid metal using water", Nuclear Engineering and Technology, 52(9), pp.1963~1973 (2020), 한국원자력학회[DOI]
- H. Hoffman, "The status of thermal-hydraulic studies on the decay heat removal by natural convection using RAMONA and NEPTUN models", Proceedings of the 4th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (1994)
- Overview report of RANONA-NEPTUN program on passive decay heat removal, No. FZKA–5667. Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany), "D. Weinberg", Inst. fuer Angewandte Thermo-und Fluiddynamik, ~ (1996)
- A. Ono, "Study on Reactor Vessel Coolability of Sodium Cooled Fast Reactor under Severe Accident Condition Water Experiments Using a Scale Model", ICAPP 2017 (2017)
- A. Bar-Cohen, "Thermally optimum spacing of vertical, natural convection cooled, parallel plates", Journal of Heat Transfer, 106, pp.116~123 (1984)
- R. A. Wirtz, "Experiments on free convection between vertical plates with symmetric heating", Journal of Heat Transfer, 104, pp.501~507 (1982)
- X. Cheng, "Turbulent natural convection coupled with thermal radiation in large vertical channels with asymmetric heating", International Journal of Heat and Mass Transfer, 41, pp.1681~1692 (1998)
- Experimental and Analytical Studies of a Passive Shutdown Heat Removal System for Advanced LMRs, "J. Heineman", Argonne National Laboratory, ~, pp.CONF-880532-10 (1988)
- S. Y. Kim, "Flow visualization experiment in a two-side wall heated rectangular duct for turbulence model assessment", Nuclear Engineering and Design, 341, pp.284~296 (2019)
- 김경문, "Design of A scale-down experimental model for SFR reactor vault cooling system performance analyses", Nuclear Engineering and Technology, 52(8), pp.1611~1625 (2020), 한국원자력학회[DOI]
- K. M. Kim, "Experimental study of turbulent air natural convection in open-ended vertical parallel plates under asymmetric heating conditions", International Journal of Heat and Mass Transfer, 159, pp.120135~ (2020)
- K. M. Kim, "Transition phenomena of natural convection of the air in an asymmetrically heated vertical channel with a damper", International Journal of Heat and Mass Transfer, 183, pp.122196~ (2022)
- G. Wu, "Assessment of RVACS performance for small size lead-cooled fast reactor", Annals of Nuclear Energy, 77, pp.310~317 (2015)
- C. Choi, "Thermal-hydraulic analyses of passive reactor vault cooling system(RVCS)in PGSFR using MARS-LMR", Annals of Nuclear Energy, 117, pp.333~342 (2018)
- S. Wang, "An experiment-based validation of a system code for prediction of passive natural circulation in sodium-cooled fast reactor", International Journal of Energy Research, 45, pp.12093~12109 (2021)
- NSSS Design and Validation of Prototype Gen-IV Sodium Cooled Fast Reactor, Mechanical Design of Prototype Gen-IV Sodium Cooled Fast Reactor, "G. H. Koo", KAERI, ~, pp.KAERI (2015), RR–4048[DOI]
