Burnup calculation of the OECD VVER-1000 LEU benchmark assembly using MCNP6 and SRAC2006
Main Article Content
Abstract
The present work aims to perform burnup calculation of the OECD VVER-1000 LEU (low enriched uranium) computational benchmark assembly using the Monte Carlo code MCNP6 and the deterministic code SRAC2006. The new depletion capability of MCNP6 was applied in the burnup calculation of the VVER-1000 LEU benchmark assembly. The OTF (on-the-fly) methodology of MCNP6, which involves high precision fitting of Doppler broadened cross sections over a wide temperature range, was utilized to handle temperature variation for heavy isotopes. The collision probability method based PIJ module of SRAC2006 was also used in this burnup calculation. The reactivity of the fuel assembly, the isotopic concentrations and the shielding effect due to the presence of the gadolinium isotopes were determined with burnup using MCNP6 and SRAC2006 in comparison with the available published benchmark data. This study is therefore expected to reveal the capabilities of MCNP6 and SRAC2006 in burnup calculation of VVER-1000 fuel assemblies.
Article Details
Keywords
VVER-1000 LEU fuel assembly, fuel burnup, MCNP6, SRAC2006
References
[2]. L. Thilagam, C. S. Sunny, V. Jagannathan, K. V. Subbaiah, “A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL”, Annals of Nuclear Energy 36, 505-519, 2009.
[3]. N. Petrov, G. Todorova, N. P. Kolev, “APOLLO2 and TRIPOLI4 solutions of the OECD VVER-1000 LEU and MOX assembly benchmark”, Annals of Nuclear Energy 55, 93–107, 2013.
[4]. M. Zheng, W. Tian, H. Wei, D. Zhang, Y. Wu, S. Qiu, G. Su, “Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment”, Annals of Nuclear Energy 63, 491–498, 2014.
[5]. L. Mercatali, A. Venturini, M. Daeubler, V. H. Sanchez, “SCALE and SERPENT solutions of the OECD VVER-1000 LEU and MOX burnup computational benchmark”, Annals of Nuclear Energy, 83, 328-341, 2015.
[6]. H. K. Louis, E. Amin, “The Effect of burnup on reactivity for VVER-1000 with MOXGD and UGD fuel assemblies using MCNPX code”, Journal of Nuclear and Particle Physics, 6 (3) 61-71, 2016.
[7]. S. A. Khan, V. Jagannathan, U. Kannan, A. Mathur, “Study of VVER-1000 OECD LEU and MOX computational benchmark with VISWAM code system”, Nuclear Energy and Technology 2, 312-334, 2016.
[8]. L. B. Vien, L. V. Vinh, H. T. Nghiem, N. K. Cuong, “Design analyses for full core conversion of the Dalat nuclear research reactor”, Nuclear Science and Technology, Vol. 4, No. 1, p.10-25, March 2014.
[9]. N. N. Dien, L. B. Vien, L. V. Vinh, H. T. Nghiem, N. K. Cuong, “Full core conversion and operational experience with LEU fuel of the Dalat nuclear research reactor”, RERTR 2014 - 35th International Meeting on Reduced Enrichment for Research and Test Reactors, IAEA Vienna Interational Center, Vienna, Austria, October 2014.
[10]. N. K. Cuong, H. T. Nghiem, V. H. Tan, “The Development of depletion program coupled with Monte Carlo computer code”, The 11th National Conference on Nuclear Science and Technology, Da Nang,Vietnam, August 2015.
[11]. N. H. Hiep, N. H. Tiep, T. V. Phu, N. T. Khai, “Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies”, Nuclear Science and Technology, Vol. 6, No. 3, pp. 16-30, 2016.
[12]. D. B. Pelowitz et al., MCNP6TM User’s Manual, Version 1.0, LA-CP-13-00634, Rev. 0, 2013.
[13]. K. Okumura, T. Kugo, K. Kaneko, and K. Tsuchihashi, SRAC2006: A Comprehensive Neutronics Calculation Code System, JAEA-Data/Code 2007-004, 2007.
[14]. F. Brown, W. Martin, G. Yesilyurt, S. Wilderman, On-The-Fly Neutron Doppler Broadening for MCNP, Los Alamos National Laboratory, LA-UR-12-00700, 2012.