SERPENT CODE POINT KINETIC PARAMETERS ANALYSIS: A NEXUS FOR NIGERIA RESEARCH REACTOR – 1 CORE CONVERSION USING ENRICHED URANIUM DIOXIDE FUEL

  • Aliyu Umar Abubakar
  • Aminu Ismaila
  • Yamusa Abdullahi Yamusa
Keywords: Delayed Neutron, NIRR-1, Point Kinetic Parameters, Serpent Code, Uranium dioxide

Abstract

The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium

References

Azande, T. S. (2011). "Nigerian Research Reactor-1 Core Neutronics Calculations", Continental Journal of Applied Sciences, 6(1): 42–46.

Azande, T. S., Balogun, G. I., Ajuji, A. S., Jonah, S. A. and Ahmed, Y. A. (2010). "The use of WIMS and CITATION codes in fuel loading required for the conversion of HEU MNSR core to LEU", Ann. Nucl. Energy, 37(9): 1223–1228.

Bergenas, J. and Scheinman, L. (2008). "The role of African regional and subregional organizations in implementing Resolution 1540", in Scheinman, L. (ed.) Implementing Resolution 1540: The Role of Regional Organizations. New York and Geneva: United Nations Institute for Disarmament Research, pp. 141–151.

Bradley, E., P. Adelfang and I. Goldman (2007). "Status and Progress of IAEA Activities on Research Reactor Conversion and Spent Fuel Return Programmes in the Years 2005-2006". Proceedings of the 29th International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR 2007).

Bretscher, M. M. (1997). Perturbation-Independent Methods For Calculating Research Reactor Kinetic Parameters. United States of America, Argonne National Laboratory, Argonne, IL 60439-4841.

Briesmeister, J. F. (2000). MCNPTM – A General Monte Carlo N-Particle Transport Code, Los Alamos National Laboratory.

Dan G. C. (2010). "Handbook of Nuclear Engineering", Springer Science & Business Media LLC, 2010.

Farhan, M. (2010). "Kinetic parameters of a low enriched uranium fuelled material test research reactor at end-of-life" Annals of Nuclear Energy 37: 1411–1414.

Fensin, M. L., Hendricks, J. S. and Mckinney, G. W. (2009). "Monte Carlo Burnup Interactive Tutorial Depletion theory Fixed passive source calculation.

Gehin, J.C., Carbajo, J.J., Ellis, R.J. (2004) "Issues in the Use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels", ORNL/TM-2004/223.

Ghasabyan, L. K. (2013). "Use of Serpent Monte-Carlo Code for Development of 3D Full-Core Models of Gen-IV Fast-Spectrum Reactors and Preparation of Group Constants for Transient Analysis with PARCS/TRACE Coupled System".

Heba K. L. (2021). "Sensitivity Analysis of the Kinetic Parameters to Physical Parameters Variation in VVER Reactor", Journal of Nuclear and Particle Physics 11(1): 15-26

Herrero, J. J., A. Vasiliev, M. Pecchia, H. Ferroukhi and S. Caruso (2016). "Review calculations for the OECD/NEA Burn-up Credit Criticality Safety Benchmark." Annals of Nuclear Energy 87: 48-57.

Hetrick, D.L. (1997). "Dynamics of Nuclear Reactors", University of Chicago Press.

Housiadas, C. (2000). Thermal-hydraulic Calculations for the GRR-1 Research Reactor Core Conversion to Low Enriched Uranium Fuel. Athens, Greece, Institute of Nuclear Technology and Radiation Protection.

Ibrahim, Y. V, Adeleye, M. O., Njinga, R. L., Odoi, H. C. and Jonah, S. A. (2015). "Prompt neutron lifetime calculations for the NIRR-1 reactor", Adv. Energy Res.,3(2): 125–131.

Jonah, S. A., Y. V. Ibrahim, A. S. Ajuji and M. Y. Onimisi (2012). "The impact of HEU to LEU conversion of commercial MNSR: Determination of neutron spectrum parameters in irradiation channels of NIRR-1 using MCNP code." Annals of Nuclear Energy 39 15–17.

Jonah, S. A. (2011). "Measured and simulated reactivity insertion transients characteristics of NIRR-1", Annals of Nuclear Energy, 38(2–3): 295–297.

Jonah, S. A. and Ahmed, Y. A. (2016). "Update on Activities Towards Conversion of NIRR-1: Progress and Challenges, in RERTR 2016" - 37th International Meeting on Reduced Enrichment for Research and Test Reactors. Radisson Blu Astrid Hotel, Antwerp, Belgium, October 23-27, 2016.

Jonah, S. A., Ibikunle, K. and Li, Y. (2009a). "A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP", Annals of Nuclear Energy, 36(8): 1285–1286.

Jonah, S. A., Ibikunle, K. and Li, Y. (2009b). "A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP", Annals of Nuclear Energy, 36: 1285–1286.

Kępisty, G., Oettingen, M., Stanisz, P., and Cetnar, J. (2017). "Statistical error propagation in HTR burnup model", Annals of Nuclear Energy. 105: 355–360.

Korkmaz, M. E. and Agar, O. (2014). "The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium-cooled fast reactor", Nuclear Engineering and Technology, 46(3): 407–412.

Lamarsh, J. R. and Baratta, A. J. (2001). "Introduction to Nuclear Engineering", Third Edition, Prentice Hall, (ISBN: 0-201-82498-1).

Leppänen, J. (2009a). "On the Use of the Continuous-Energy Monte Carlo Method for Lattice Physics Applications", in 2009 International Nuclear Atlantic Conference - INAC 2009. Rio de Janeiro,RJ, Brazil, September 27 to October 2, 2009.

Leppänen, J. (2009b). "Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation", Annals of Nuclear Energy, 36(7): 878–885.

Leppänen, J., M. Aufiero, E. Fridman, R. Rachamin and S. van der Marck (2014). "Calculation of effective point kinetics parameters in the Serpent 2 Monte Carlo code." Annals of Nuclear Energy, 65: 272-279.

Leppänen, J., M. Pusa, T. Viitanen, V. Valtavirta and T. Kaltiaisenaho (2015). "The Serpent Monte Carlo code: Status, development and applications in 2013." Annals of Nuclear Energy 82: 142-150.

Leppänen, J. and Viitanen, T. (2012). Serpent Monte Carlo Neutron Transport Code NEA Expert Group on Advanced Monte Carlo Techniques, Meeting. Finland. Available at: https://www.oecd-nea.org/science/wpncs/amct/workingarea/meeting2012/Serpent-EGAMC-2012-09-17.pdf (Accessed: 3 September 2018).

Luka, S., K. Andrej, Ž. Gašper and R. Matjaž (2008). "Monte Carlo Calculation of Kinetic Parameters for the TRIGA Mark II Research Reactor". Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia.

Meggitt, G. (2006). "Fission, critical mass and safety - A historical review, " J. Radiol. Prot., 26(2): 141–159.

Mengjiao, W., L. Yiguo, W. Xiaobo, P. Dan, H. Jingyan, H. Qian, Z. Jinhua and L. Jin (2017). A New Low Enrichment Uranium Core Design of MNSR. 2017 25th International Conference on Nuclear Engineering.

Muhammad, F. and A. Majid (2008). "Effects of high-density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor." ANUCENE 35: 1720–1731.

National Nuclear Security Administration (NISA) (2011). Criticality Safety Qualification Standard Reference Guide.

Odoi, H. C. and Gbadago, J. K. (2017). "Conversion of the Ghana’s Miniature Neutron Source Reactor from HEU to LEU" : 2017 Report’, pp. 1–5.

Ott, K.O. and Neuhold, R.J. (1985). "Introductory Nuclear Reactor Dynamics", American Nuclear Society, La Grange Park, IL, USA.

Ramos, M., Ferrer, S. and Villaescusa, J. (2004). "Digitalization of a mammographic phantom view through a Monte Carlo simulation", 11th International Congress of the International Radiation Protection Association, pp. 1–10. Available at: http://pintassilgo2.ipen.br/biblioteca/cd/irpa/2004/files/4f3.pdf.

Salawu, A. (2020). A new 3D Reactor model for criticality safety analysis of NIRR-1 using Scale 6.2.3. FUDMA journal of science, vol. 4 N0. 1 pp 401-405.

Stacey, W. M. (2018). Nuclear Reactor Physics. Third, Rev. Weinheim,Germany: Wiley-VCH.

Stacey, W. M. (2007). Nuclear Reactor Physics. Second Edi.,WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Svetozar, M., Jan, H., and Gabriel, F. (2008). "MCNP5 Delayed Neutron Fraction (βeff) Calculation in Training Reactor VR–1".,Journal of Electrical Engineering 59 (4): 221-224.

Woodruff, W. L. and J. R. Deen (1994). Transient analyses and thermal-hydraulic safety margins for the Greek Research Reactor (GRRI). International Meeting on Reduced Enrichment for Research and Test Reactors Williamsburg, Virginia, USA, US Department of Energy, Office of Nonproliferation and National Security.
Published
2022-06-11
How to Cite
Abubakar, A. U., Ismaila, A., & Yamusa, Y. A. (2022). SERPENT CODE POINT KINETIC PARAMETERS ANALYSIS: A NEXUS FOR NIGERIA RESEARCH REACTOR – 1 CORE CONVERSION USING ENRICHED URANIUM DIOXIDE FUEL. FUDMA JOURNAL OF SCIENCES, 6(2), 203 - 207. https://doi.org/10.33003/fjs-2022-0602-954