Verification of DeCART2D/CAPP code system for VHTR analysis with PMR-200 benchmark Eun Jeong a , Jinsu Park a , Hyun Chul Lee b , Peng Zhang a , Jiankai Yu a , Matthieu Lemaire a , Sooyoung Choi a , Deokjung Lee a,⇑ a Department of Nuclear Engineering Ulsan National Institute of Science and Technology (UNIST), Ulsan 44919, Republic of Korea b Mechanical Engineering, Pusan National University (PNU), Busan, Republic of Korea article info Article history: Received 28 March 2018 Received in revised form 3 April 2019 Accepted 2 May 2019 Keywords: Generation-IV reactor VHTR PMR-200 DeCART2D/CAPP McCARD abstract This paper presents the verification of the DeCART2D/CAPP code system for the Very High Temperature Gas-Cooled Reactor (VHTR) analysis with the Prismatic Modular Reactor 200 (PMR-200) benchmark. The McCARD Monte Carlo (MC) code is used to obtain the reference solution. The verification has been per- formed for the effective multiplication factor (k eff ) and reactivity coefficients at the levels of fuel compact, fuel block, and full core. Furthermore, the verification of the depletion calculation has been conducted for the fuel block and the verification for the power distribution has been performed at the levels of fuel block, two-dimensional (2D) and three-dimensional (3D) full core. The verification results of DeCART2D, CAPP, and DeCART2D/CAPP are compared systematically against the reference McCARD solu- tions to demonstrate the VHTR modeling capability and accuracy of the codes. It was successfully shown that the k eff errors of the DeCART2D/CAPP code system are smaller than 510 pcm, the isothermal tem- perature coefficient (ITC) errors are smaller than 0.66 pcm/K, and the power distribution errors are smaller than 2.80%. It was also shown that the maximum k eff errors of DeCART2D fuel block depletion cal- culations are smaller than 460 pcm. Ó 2019 Published by Elsevier Ltd. 1. Introduction Generation IV reactors are a set of nuclear reactors currently under research for commercial applications by the Generation IV International Forum, with Technology readiness levels varying between the level requiring a demonstration to economical com- petitive implementation (Locatelli et al., 2013). The Generation- IV reactor candidates are Molten Salt Reactor (MSR), Sodium- cooled Fast Reactor (SFR), VHTR, Lead-cooled Fast Reactor (LFR), Gas-cooled Fast Reactor (GFR), and Super Critical Water-cooled Reactor (SCWR). The research on these designs is continuously pro- ceeding in an active manner (Park et al., 2015; Tak et al., 2015). The VHTR uses a graphite moderator and a helium coolant (Moses, 2010). The use of these materials makes it possible for VHTRs to safely produce high coolant outlet temperatures in the range of 1000  1300 K (Chapin et al., 2004). Such high coolant outlet tem- peratures allow high efficiency in electric power generation as well as hydrogen production and other industrial process-heat applica- tions. The development of advanced simulation tools for the VHTR core is an essential research field for the development of next gen- eration reactors. Korea Atomic Energy Research Institute (KAERI) has chosen a block-type VHTR composed of a complicated fuel structure within the fuel block and an annular fuel-reflector arrangement (Joo et al., 2006). KAERI has developed DeCART2D code for VHTR analysis and this paper presents the verification of the code for the block-type VHTR. Argonne National Laboratory (ANL) used a lattice transport code DRAGON and a core analysis code DIF3D to analyze High Temper- ature Engineering Test Reactor (HTTR) (Taiwo et al., 2005; Taiwo and Kim, 2006). In that study, the error of k eff was reported to be approximately 1700 pcm compared to the result of Monte-Carlo code. KAERI used a lattice transport code HELIOS and a core anal- ysis code MASTER for VHTR analysis (Noh et al., 2008). In this HELIOS/MASTER analysis, Reactivity-equivalent Physical Transfor- mation (RPT) method was applied to treat the double heterogene- ity of the tri-structural-isotropic (TRISO) fuel particles dispersed in the fuel compact (Kim et al., 2017). In the HTTR analysis of Idaho https://doi.org/10.1016/j.anucene.2019.05.002 0306-4549/Ó 2019 Published by Elsevier Ltd. ⇑ Corresponding author at: Department of Nuclear Engineering Ulsan National Institute of Science and Technology, 50 UNIST-gil, Ulsan 44919, Republic of Korea. E-mail addresses: eunjeong@unist.ac.kr (E. Jeong), jinsu@unist.ac.kr (J. Park), hyunchul.lee@pusan.ac.kr (H.C. Lee), zhangpeng@unist.ac.kr (P. Zhang), jackyue@ unist.ac.kr (J. Yu), mlemaire@unist.ac.kr (M. Lemaire), schoi@unist.ac.kr (S. Choi), deokjung@unist.ac.kr (D. Lee). Annals of Nuclear Energy 133 (2019) 154–168 Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene