Advances in Energy Research

Techno-Press (테크노프레스)
권호 표지
DOI QR코드

DOI QR Code

Prompt neutron lifetime calculations for the NIRR-1 reactor

  • Ibrahim, Yakubu V. (Centre for Energy Research and Training, Ahmadu Bello University) ;
  • Adeleye, Micheal O. (Bingham University) ;
  • Njinga, Raymond L. (Ibrahim Badamasi Babangida University) ;
  • Odoi, Henry C. (School of Nuclear and Allied Sciences, University of Ghana) ;
  • Jonah, Sunday A. (Centre for Energy Research and Training, Ahmadu Bello University)
  • Received : 2015.03.04
  • Accepted : 2015.06.30
  • Published : 2015.06.25
⟨ Previous Next ⟩

Abstract

Prompt neutron lifetime calculations have been performed for the NIRR-1 reactor HEU and LEU cores using the 1/v insertion and the Adjoint flux weighing methods. Results of calculations obtained for the HEU and LEU cores are respectively $57.3{\pm}0.8$ and $47.5{\pm}0.7$ for the 1/v insertion and $56.9{\pm}0.3$ and $46.3{\pm}0.5$ for the Adjoint flux. There is a good agreement seen between the two methods for both cores. The prompt neutron lifetime was observed to be shorter in the LEU than for the HEU as expected. However, the Adjoint flux weighing method seemed to be the easiest method in calculating the prompt neutron lifetime for NIRR-1.

Keywords

References

  1. Abrefah, R.G., Odoi, H.C., Ampomah-Amoako, E. and Mensimah, E. (2012), "Investigation into the effects of Zircaloy-4 on the neutronics of low enriched uranium MNSR core", Prog. Nucl. Energy, 54(1), 5-10. https://doi.org/10.1016/j.pnucene.2011.09.009
  2. Abtin, F., Feghhi, S.A.H. and Jafarikia, S. (2013), "Neutronic evaluations for MNSR research reactor core conversion from HEU to LEU", Ann. Nucl. Energy, 51, 69-73. https://doi.org/10.1016/j.anucene.2012.08.016
  3. Albert, L.H. and Diamond, D.J. (2006), "Calculation of the prompt neutron lifetime in NBSR", Nucl. Sci. Eng., 153, 26-32. https://doi.org/10.13182/NSE06-4
  4. Bretscher, M.M. (1997), "Perterbation-Independent Methods for Calculating Research Reactor Kinetic Parameters", ANL/RERTR/TM-30, Argonne national Laboratory, December.
  5. Final Safety Analysis Report of Nigeria Research Reactor-1, (2005), CERT Technical Report-CERT/NIRR-1/FSAR-01.
  6. Haddad, K. and Boush, M. (2011), "Validation of MCNP volume efficiency calculation for gamma spectrometric assay of large NORM samples", J. Radio. Nucl. Chem., 289, 97-101. https://doi.org/10.1007/s10967-011-1048-4
  7. Hamidatou, L. and Benkharfia, H. (2011), "Experimental and MCNP calculations of neutron flux parameters in irradiation channel at Es-Salam reactor", J. Radio. Nucl. Chem., 287, 971-975. https://doi.org/10.1007/s10967-010-0922-9
  8. Heba, K.L. (2014), "Evaluation of effective prompt neutron lifetime for pressurized water reactors", Nucl. Tech., 188, 1-7. https://doi.org/10.13182/NT13-120
  9. Kiedrowski, B., Brown, F. and Wilson, P. (2011), "Adjoint-weighted tallies for k-eigenvalue calculations with continuous-energy Monte Carlo", Nucl. Sci. Eng., 168, 226-241. https://doi.org/10.13182/NSE10-22
  10. Odoi, H.C., Akaho, E.H.K., Jonah, S.A., Abrefah, R.G. and Ibrahim, V.Y. (2014), "Study of criticality safety and neutronic performance for a 348-Fuel-Pin Ghana research reactor-1 LEU core using MCNP code", World J. Nucl. Sci. Tech., 4, 46-52. https://doi.org/10.4236/wjnst.2014.41008
  11. Ibrahim, Y.V., Odoi, H.C., Njinga, R.L., Odeleye, M.O. and Jonah, S.A. (2012), "Monte Carlo simulation of additional safety control rod for commercial MNSR to enhance safety", Ann. Nucl. Energy, 44, 71-75. https://doi.org/10.1016/j.anucene.2012.01.008
  12. Ibrahim, Y.V., Odoi, H.C., Thomas, J.W. and Jonah, S.A. (2013), "Design options of control rod for low enriched uranium fueled NIRR-1 using Monte Carlo N-particle code", J. Nucl. Energy Sci. Power Gen. Tech., 2, doi:10.4172/2325-9809.1000104.
  13. Jonah, S.A., Liaw, J.R. and Matos, J.E. (2007), "Monte Carlo simulation of core physics parameters of nigerian research reactor-1", Ann. Nucl. Energy, 34, 953-957. https://doi.org/10.1016/j.anucene.2007.05.010
  14. Jonah, S.A., Ibikunle, K. and Li, Y. (2009), "A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP", Ann. Nucl. Energy, 36, 1285-1286. https://doi.org/10.1016/j.anucene.2009.05.001
  15. Jonah, S.A., Ibrahim, Y.V., Ajujia, A.S. and Onimisi, M.Y. (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", Ann. Nucl. Energy, 39, 15-17. https://doi.org/10.1016/j.anucene.2011.08.026
  16. Khattab, K. and Sulieman, I. (2009), "Calculations of the thermal and fast neutron fluxes in the Syrian MNSR irradiation sites using the MCNP-4C code", Appl. Radiat. Isotop., 67, 535-538. https://doi.org/10.1016/j.apradiso.2008.11.002
  17. Li, Y., Ke, G., Sun, Z., Shen, F., Liu, T. and Zhou, Y. (2009), "The study of physics and thermal characteristics for in-hospital neutron irradiator (INHI)", Appl. Radiat. Isotop., 67, S234-S237. https://doi.org/10.1016/j.apradiso.2009.03.117
  18. MCNP5 (2003), A general Monte Carlo N-Particle Code, LA-UR-03-1987, Updates to the MCNP5 used for the delayed neutron fraction and prompt neutron lifetime calculations, Eds. B.C. Kiedrowski, T.E. Booth, F.B. Brown, J.S. Bull, J.A. Favorite, R.A. Forster, R.L. Martz, MCNP5-1.60 Feature Enhancements & Manual Clarifications, LA-UR-10-06217.
  19. Peters, N.J., Brockman, J.D. and Robertson, J.D. (2009), "Using Monte Carlo transport to accurately predict isotope production and activation analysis rates at the University of Missouri research reactor", J. Radio. Nucl. Chem., 282, 255-259. https://doi.org/10.1007/s10967-009-0163-y
  20. Rabira, M.H. and Haha, C.J. (2013), "Evaluation of Delayed Neutron Fraction of RTP using MCNP", Transactions of the Korean Nuclear Society Spring Meeting, Gwangju, Korea, May.
  21. Shaaban, I. (2010), "Design of the thermal neutron beam for neutron radiography at the Syrian MNSR", Ann. Nucl. Energy, 37, 1588-1594. https://doi.org/10.1016/j.anucene.2010.06.003
  22. Snoj, L., Kavcic, A., Zerovnik, G. and Ravnik, M. (2010), "Calculation of kinetic parameters for mixed TRIGA cores with Monte Carlo", Ann. Nucl. Energy, 37, 223-229. https://doi.org/10.1016/j.anucene.2009.10.020
  23. Travelli, F.A. (1999), Progress of RERTR Programme, 22nd RERTR Meeting Budapest, Hungary.
  24. Verboomen, B., Haec, W. and Baeten, P. (2006), "Monte Carlo calculation of the effective neutron generation time", Annal. Nucl. Energy, 33, 911-916. https://doi.org/10.1016/j.anucene.2006.05.001

Cited by

  1. Study of fission gas products effect on thermal hydraulics of the WWER1000 with enhanced subchannel method vol.5, pp.2, 2015, https://doi.org/10.12989/eri.2017.5.2.091
  2. Sensitivity analysis of numerical schemes in natural cooling flows for low power research reactors vol.5, pp.3, 2017, https://doi.org/10.12989/eri.2017.5.3.255