Detailed neutronic analysis of a MOX-fueled metal-cooled reactor Zeeshan Jamil a,b , Bin Wu a , Shengpeng Yu a , Qi Yang a , Muhammad Salman Khan a,b , Muhammad Younas Ali a,b , Liqin Hu a,c,⇑ a Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031, China b University of Science and Technology of China, Hefei, Anhui 230027, China c Collaborative Innovation Center of Radiation Medicine of Jiangsu Higher Education Institutions, Su Zhou 215006, China article info Article history: Received 3 August 2017 Received in revised form 29 November 2017 Accepted 1 December 2017 Keywords: SuperMC BFS-62-3A Stainless steel reflector Code-to-code verification Sodium void reactivity effect abstract With the prominent features of different U-Pu fuel compositions, the MOX-fueled fast spectrum metal- cooled reactors have been of increasing interest for recent years worldwide. Present study invokes a detailed investigation on the calculations for neutronic analysis and it conducts a comparative study for neutron physics parameters of a metal-cooled fast spectrum reactor – the BFS-62-3A. This work is an amalgamation of four tasks. In the first part, a detailed comparative analysis was performed to perform a code-to-code verification using the available results of three Monte Carlo codes including SuperMC, MCNP, and Serpent, and a deterministic one, the DYN3D-MG with the employment of continuous neutron energy cross-sections. The experimental results of BFS-62-3A benchmark were used to assess the poten- tiality of the aforementioned reactor physics codes. For most of the integral parameters, the SuperMC was found to be on the leading edge. In the second part, the effects of data libraries including ENDF/B-7.1, ENDF/B-7.0, ENDF/B-6.6, JEFF3.2, and HENDL3.0 on the simulations performed using SuperMC code for evaluating k-eff and radial fission rates were all assessed. The third part incorporates the investigation of the fission rates’ deviations in stainless steel radial reflector by changing the density of the reflector. The decrease of density by 5% was found to be in good agreement with the benchmark. In the last part, that has a special importance to the concept pertaining to safety-enhanced Sodium-cooled Fast Reactor (SFR) core, the reactivity of the critical assembly was studied by calculating the sodium void reactivity effect. The simulation results of SuperMC code agreed well with the available experimental and simula- tion results. The present study has enabled SuperMC code to pass another big milestone on dealing with complex and advanced nuclear systems. Ó 2017 Elsevier Ltd. All rights reserved. 1. Introduction At present, the nuclear power from fission sources contributes a share of over 11% of the world’s available electricity generating resources and an increasing tendency could lead that up to 17% by 2050 (Wu et al., 2016; A Survey Report, 2013). Progressing nuclear energy sector demands for the R&D to be made that should be in line with engineering and technology of the advanced nuclear systems, the Gen IV concepts, for instance (Pioro, 2016). An SFR, BFS-62-3A, has been under employment to be investigated by var- ious researchers because of its dominating features that might include its multi-zoned MOX-fueled structure, etc. The aforemen- tioned test reactor along with many others has been maintained at Big Physical Facility (BFS2) at IPPE – A Russian institute. Many notable experiments have been conducted overs there with critical facilities by the nuclear professionals to pave a way to benchmark- ing and to further perform the detailed analysis of results obtained through the experimental and simulation studies (Dulin et al., 2014; Mitenkova et al., 2013; Manturov et al., 2006; Oberkampf and Trucano, 2006; Hazama et al., 2004). Furthering the research on metal-cooled fast spectrum reactors, a lot still needs to be done. Beside the present work, the BFS-62-3A benchmark was employed by us in our recent work that was con- ducted over here at INEST, CAS China in which the Monte Carlo code, Super Monte Carlo Program for Nuclear and Radiation Simu- lation (SuperMC), was validated against the benchmark (BFS-62- 3A). A number of neutron physics parameters were calculated. The performed work generated a good reason to further investigate the simulated reactor’s model. Different Monte Carlo and deter- ministic codes with different nuclear data libraries have been used https://doi.org/10.1016/j.anucene.2017.12.004 0306-4549/Ó 2017 Elsevier Ltd. All rights reserved. ⇑ Corresponding author at: Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031, China. E-mail address: liqin.hu@fds.org.cn (L. Hu). Annals of Nuclear Energy 114 (2018) 427–436 Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene