Code validation of a scaled-down DRACS model in RELAP5/SCDAPSIM/MOD 4.0

Journal Article · Sat Aug 18 00:00:00 EDT 2018 · Annals of Nuclear Energy (Oxford)

DOI:https://doi.org/10.1016/j.anucene.2018.05.058· OSTI ID:1504426

Lin, Hsun-Chia ^[1]; Lv, Qiuping ^[2]; Shi, Shanbin ^[1]; Sun, Xiaodong ^[1]; Christensen, Richard ^[3]; Yoder, Graydon ^[4]

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1. The Ohio State Univ., Columbus, OH (United States); Univ. of Michigan, Ann Arbor, MI (United States)
2. The Ohio State Univ., Columbus, OH (United States); Argonne National Lab. (ANL), Lemont, IL (United States)
3. The Ohio State Univ., Columbus, OH (United States); University of Idaho, Idaho Falls, ID (United States)
4. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs). Additionally, benchmark study of decay heat removal capability of DRACS by natural circulation is critical for FHR safety analysis. Two scaled-down test facilities, a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF), were designed following a scaling analysis and constructed to investigate thermal performance and heat removal capability of the DRACS. The LTDF uses water as a surrogate coolant for both the primary and DRACS loops. Two transient scenarios were carried out in the LTDF, namely, DRACS startup and primary coolant pump trip. In both of the scenarios, the experimental results demonstrate sufficient heat removal capabilities of the DRACS with natural circulations established in the system. These two scenarios were simulated using RELAP5/SCDASIM/MOD4.0 code and the code simulation results show good agreement with the LTDF experimental data. For the HTDF, molten salts FLiNaK and KF-ZrF4 are used respectively as the primary coolant and DRACS coolant. Thermodynamic and transport properties of FLiNaK and KF-ZrF4 were implemented into RELAP5/SCDAPSIM/MOD 4.0. Lastly, simulations of transient scenarios in the HTDF were also performed and the code results will be compared with the HTDF experimental data once available

• Research Organization: Argonne National Laboratory (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE). Nuclear Energy University Programs (NEUP)
• Grant/Contract Number: AC02-06CH11357
• OSTI ID: 1504426
• Alternate ID(s): OSTI ID: 22834136; OSTI ID: 1692221
• Journal Information: Annals of Nuclear Energy (Oxford), Journal Name: Annals of Nuclear Energy (Oxford) Journal Issue: C Vol. 121; ISSN 0306-4549
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
AHTR
Code validation
DRACS
FHR
Natural circulation
RELAP5