Article Sidebar

PDF
Published: 2016-09-30
Abstract Views: 148
Views PDF: 62
DOI: https://doi.org/10.53747/jnst.v6i3.160
Issue
Vol. 6 No. 3 (2016)
Section
Articles
How to Cite
Nguyen , H. H., Nguyen , H. T., Tran , V. P., & Nguyen , T. K. (2016). Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies. Nuclear Science and Technology, 6(3), 16-30. https://doi.org/10.53747/jnst.v6i3.160

Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies

Nguyen Huy Hiep1, Nguyen Huu Tiep1, Tran Viet Phu1, Nguyen Tuan Khai1
1 Nuclear Power Center, Institute for Nuclear Science and Technology 179 Hoang Quoc Viet, Nghia Do, Cau Giay, Hanoi, Vietnam

Main Article Content

Abstract

The paper aims to develop an MCNP5-ORIGEN2 coupling scheme for burnup calculation. Specifically, the Monte Carlo neutron transport code (MCNP5) and the nuclides depletion and decay calculation code (ORIGEN2) are combined by data processing and linking files written in the PERL programming language. The validity and applicability of the developed coupling scheme are tested through predicting the neutronic and isotopic behavior of the “VVER-1000 LEU Assembly Computational Benchmark”. The MCNP5-ORIGEN2 coupling results showed a good agreement with the k-inf benchmark values within 600 pcm during the entire burnup history. In addition, the differences of isotopes concentration at the end of the burnup (40 MWd/kgHM) when compared with benchmark values were reasonable and generally within 6.5%. The developed coupling scheme also considered the shielding effect due to gadolinium isotopes and simulated well the depletion of isotopes as a function of the radial position in gadolinium bearing fuel rods.

Article Details

Keywords

MCNP5, ORIGEN2, burnup, coupling, VVER-1000 LEU

References

[1] Keisuke Okumura, Yasunobu Nagaya, Takamasa Mori, “MVP-BURN: Burn-up Calculation Code Using a Continuos-energy Monte Carlo Code MVP”, Japan Atomic Energy Agency (JAEA), January 2005.
[2] Edward J. Parma, “BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies”, Sandia National Laboratories, November 2002.
[3] X-5 Monte Carlo Team, “MCNP – A General N-Particle Transport Code, Version 5”, Los Alamos National Laboratory, April 2003.
[4] A. G. Croff, “A User’s Manual for the ORIGEN2 Computer Code”, Oak Ridge National Laboratory, July 1980.
[5] R.L. Moore, B. G. Schnitzler, C. A. Wemple, R. S. Babcock, D. E. Wessol, “MOCUP: MCNP-ORIGEN2 Coupled Utility Program,” INEL-95/0523, Idaho National Engineering Laboratory, September 1995.
[6] David I. Poston, Holly R. Trellue, “User’s Manual, Version 2.0, for Monteburns, Version 1.0”, LA-UR-99-4999, Los Alamos National Laboratory, September 1999.
[7] Zhiwen Xu, Pavel Hejzlar, Michael J. Driscoll, and Mujid S. Kazimi, “An improved MCNP-ORIGEN Depletion Program (MCODE) and its Verification for High-burnup Applications”, PHYSOR, Seoul, Korea, October 2002.
[8] F. Alvarez-Velarde, E. M. Gonzalez-Romero, I. Merino Rodriguez, “Validation of the Burn-up code EVOLCODE 2.0 with PWR Experimental Data and with a Sensitivity/Uncertainty Analysis”, Annals of Nuclear Energy, June 2014.
[9] Javier Sanz, Oscar Cabellos, Nuria García-Herranz, “ACAB Inventory Code for Nuclear Application: User’s Manual version 2008”, December 2008.
[10] Meiyin Zheng, Wenxi Tian, Hongyang Wei, Dalin Zhang, Yingwi Wu, Suizheng Qiu, Guanghui Su, “Development of an MCNP-ORIGEN Burn-up Calculation Code System and its Accuracy Assessment”, Annals of Nuclear Energy, August 2013.
[11] Luong Ba Vien, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien 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.
[12] Nguyen Nhi Dien, Luong Ba Vien, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien 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.
[13] Nguyen Kien Cuong, Huynh Ton Nghiem, Vuong Huu 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.
[14] NEA/NSC/DOC 10, “A VVER-1000 LEU and MOX Assembly Computational Benchmark”, Nuclear Energy Agency, Organization for Economic Co-operation and Development, 2002.
[15] Luka Snoj, Matiaz Ravnik, “Calculation of Power Density with MCNP in TRIGA reactor”, International Conference, Nuclear Energy for New Europe 2006, September 2006.
[16] R. E. MacFarlane, A. C. Kahler, “Methods for Processing ENDF/B-VII with NJOY”, Nuclear Data Sheets, October 2010.
[17] Go Chiba, Masashi Tsuji, Tadashi Narabayashi, Yasunori Ohoka & Tadashi Ushio, “Important Fission Product Nuclides Identification Method for Simplified Burnup Chain Construction”, Journal of Nuclear Science and Technology, 2015.