Determination of out-core fuel burnup in TRIGA PUSPATI


  • Mohamad Annuar Assadat Husain Universiti Teknologi Malaysia
  • Suhairul Hashim Universiti Teknologi Malaysia
  • Sib Krishna Ghoshal Universiti Teknologi Malaysia
  • Mohamad Hairie Rabir Malaysia Nuclear Agency
  • Norasalwa Zakaria Malaysia Nuclear Agency
  • Muhammad Rawi Mohamed Zin Malaysia Nuclear Agency
  • David Bradley University of Surrey



Neutron Flux, Thermal Neutron, Fast Neutron, MCNPX, RTP


The investigation was conducted on the out-core neutron flux and burn-up at irradiated fuel stored in TRIGA PUSPATI research reactor tank. This is required to examine whether the thermal and/or fast neutron flux can influence burn-up of the irradiated fuel stored in the same vicinity of the reactor core, the fuel rack being located 1 m above the core. MCNPX code was used to simulate fast and thermal neutron flux for the reactor operating at 750 kW. In this work, the computational model was created using MCNPX version 2.7 with the evaluated nuclear data file for thermal neutron scattering law data (ENDF7) cross-section data library and using a 10 cm x 10 cm x 10 cm mesh model. The results showed that the axial distribution for thermal neutrons occurred at energy lower than 1 x 10-6 MeV. Thermal neutron traveled at the maximum distance of 78 cm due to thermalization by moderator. Based on the maximum distance traveled by the thermal neutron, the thermal neutron did not reach the storage rack located 1 m from the core, hence there was no burn-up occurring at the irradiated fuel since burn-up can only occur in the thermal neutron region. For fast neutron, the axial distribution energy was higher than 1 x 10-6 MeV and traveled more than 158 cm. The reaction time for the fast neutron was too short to result in burn-up due to its fast travel.

Author Biographies

Suhairul Hashim, Universiti Teknologi Malaysia

 Department of Physics, Faculty of Science

Sib Krishna Ghoshal, Universiti Teknologi Malaysia

 Department of Physics, Faculty of Science

David Bradley, University of Surrey

Department of Physics, Faculty of Engineering and Physical Sciences


Alnour, I. A, Wagiran, H., Ibrahim, N., Hamzah, S., Wee, B. S., Elias, M. S., Karim, J. A. (2013) Determination of neutron flux parameters in PUSPATI TRIGA mark II research reactor. Malaysia. Journal Radioanalytical Nuclear Chemistry, 296(3), 1231-1237.

El Bakkari, B., El Bardouni, T., Nacir, B., El Younoussi, C., Boulaich Y. Boukhal, H., Zoubair, M. (2013). Fuel burn up analysis for the Moroccan TRIGA research reactor. Annals of Nuclear Energy, 51, 112-119.

Dalle, H. M., Jeraj, R., Tambourgi, E. (2015). Characterization of burned fuel of the TRIGA IPR – R1 research reactor using Monteburns code. Retrieved from

Liew, H. F. (2010). The absolute method of neutron activation analysis using TRIGA neutron reactor. M.Sc. dissertation, Universiti Teknologi Malaysia.

Mitsui, J. and Sugiyama, K. (1973). Neutron thermalization in graphite. Journal of Nuclear Science and Technology, 10(1), 1-9.

Hendricks, J. S., McKinney, G. W., Fensin, M. L., James, M. R., Johns, R. C., Durkee, J. W., Finch, J. P., Pelowitz, D. B., Waters, L. S., Johnson, M. W., Gallmeier, F. X. (2008). MCNPXTM User’s Manual. Version 2.60. Retrieved from

Rabir, M. H, Usang, M. D., Hamzah, N. S., Karim, J. A. and Salleh, M. A. S. (2013). Determination of neutronic parameters of RTP core using MCNP code. Jurnal Sains Nuklear Malaysia, 25(1), 46-60.

Rabir, M. H. and Usang, M. D. (2015). Neutron flux and power in RTP Core-15. AIP Conference Proceeding, 1706, 050018.

Snoj, L., Trkov, A., Jac´imovic, R., Rogana, P., Zerovnik, G., Ravnik, M., (2011). Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization. Applied Radiation and Isotopes, 69(1), 136–141.

Ueda, M., Kikuchi, S., Kikuchi, T., Kumanomido, H., Seino, T. (1993). Basic studies on neutron emission-rate method for burnup measurement of spent light-water-reactor fuel bundle. Journal of Nuclear Science and Technology, 30(1), 48-59.