Safety assessment and management of spent nuclear fuel for TRIGA mark II research reactor
-
Beya Heritier
und
Alya Badawi
Abstract
Democratic Republic of Congo (DRC) has a TRIGA mark II research reactor called TRICO II, its design power is 1.00 MW. The reactor was in extended shutdown state since November 2004. The DRC government has decided to resume its operation using the last uploaded core. One of the safety features to be determined before putting the spent fuel into the reactor core is the calculation of its excess reactivity, radionuclide inventories as well as its discharge burn-up. The spent fuel was modeled and simulated by using Monte Carlo software, MCNPX code. The input data and the horizontal and vertical modeling for the fuel pins, control rods and moderator were done. The model results were validated by calculating the effective delayed neutron fraction (β eff) and the worth of the control rods. The results of the criticality and fuel burn-up were compared with the reference design parameters and with the experimental measurements and it were found in good agreement. The calculations showed that the last uploaded core has 47.00 g of 235U which represent only 2% of fissile materials. The depletion analysis results showed that the highest radio-activities come from 151Sm, 137Cs, 90Y, 90Sr and 85Kr.
Acknowledgments
The authors would like to thank and acknowledge the IAEA, Nuclear and Radiation Engineering Department, Faculty of Engineering, Alexandria University and Egyptian atomic Energy Authority (EAEA) for the organization and management the AFRA program for master’s degree in Nuclear Science and Technology. Also, the authors would like to express and acknowledge the Kinshasa Nuclear Research Centre (CREN-K) and all the staff of TRICO II research reactor for providing us with all the necessary information related to the TRICO II reactor.
-
Author contributions: All the authors have accepted responsibility for the entire content of this submitted manuscript and approved submission.
-
Research funding: None declared.
-
Conflict of interest statement: The authors declare no conflicts of interest regarding this article.
References
Alyson, K. (2012). Determination of a calculation bias in the MCNP model of the OSTR, Master thesis. Oregon State University, Oregon, USA.Suche in Google Scholar
Antti, R. and Kotiluoto, P. (2016). FIR1TRIGA activity inventories for decommissioning planning, 194. VTT Technical Research Centre of Finland Ltd, pp. 28–38.10.13182/NT15-86Suche in Google Scholar
Böck, H. and Khan, R. (2010). Neutronics analysis of the TRIGA Vienna mixed core. Vienna, Austria, Vienna University of Technology/Atominstitute stadionallee 2.Suche in Google Scholar
Beya, H. (2019). Safety assessment and cost estimates for the decommissioning of D. R. Congo research reactor (TRICO II), M.Sc. Thesis dissertation, Alexandria, Egypt, Nuclear & Radiation Eng. Dept., Fac. Eng., Alex. Univ.Suche in Google Scholar
Borio di Tigliole, A., Cammi, A., Clemenza, M., Memoli, V., Pattavina, L., and Previtali, E. (2010). Benchmark evaluation of reactor critical parameters and neutron fluxes distributions at zero power for the TRIGA Mark II reactor of the University of pavia using the Monte Carlo code MCNP. Prog. Nucl. Energy 52: 494–502, https://doi.org/10.1016/j.pnucene.2009.11.002.Suche in Google Scholar
Cren, K. (2008). Logbook of TRICO II operating, Kinshasa, Centre Regional Etudes Nucl. Kinshasa (CREN-K).Suche in Google Scholar
Davide, C. (2013). Development and experimental validation of a Monte Carlo simulation model for the TRIGA mark II reactor, Ph.D. Thesis. University adeglistudi Milano-Bicocca, Milan, Italy.Suche in Google Scholar
Goorley, T. (2004). Criticality calculations with MCNP5: A primer, 2nd ed. USA: Los Alamos National Laboratory, X-5, LA-UR-04-0294.Suche in Google Scholar
Haeck, W. and Verboomen, B. (2007). An optimum approach to Monte Carlo burn-up. Nucl. Sci. Eng. 156: 180–196, https://doi.org/10.13182/NSE07-A2695.Suche in Google Scholar
Huda, M., Suaiya, J., and Obara, T. (2008). Burn-up analysis and in-core fuel management study of the 3 MW TRIGA Mark II research reactor. Annals of Nuclear Energy 35: 141–147, https://doi.org/10.1016/j.anucene.2007.05.013.Suche in Google Scholar
Monte Carlo Team (2003). MCNP – a general Monte Carlo N-particle transport code, overview and theory. UAS, Los Alamos National Laboratory.Suche in Google Scholar
Mizanur, R., Rahman, A., Hossain, S., Das, P.K., Ali, A., Soner, A.M., and Abdullah-Al-Mahmud (2021). Calculation of fuel burn-up and excess reactivity using TRIGLAV code for BAEC TRIGA research reactor. Nucl. Energy Sci Tech 14: 291–301, https://doi.org/10.1504/IJNEST.2020.117699.Suche in Google Scholar
Mohammed, M., Abdelouahed, C., and Abdelaziz, D. (2015). Calculation of kinetic parameters of the Moroccan TRIGA mark-II reactor using the Monte Carlo code MCNP. Adv. Appl. Phy 3: 1–8, https://doi.org/10.12988/AAP.2015.531.Suche in Google Scholar
Khattak, M.A., Borhana, A.A., Yasin, N., Khan, R., and Saad, J. (2018). Initial criticality analysis of Malaysia TRIGA research reactor using TRIGLAV computer code. Eng. Technol. 735: 899–903, https://doi.org/10.14419/ijet.v7i4.35.26279.Suche in Google Scholar
Malu, K. (1976). Security rapport, Kinshasa, Centre Regional Etudes Nucl. Kinshasa (CREN-K).Suche in Google Scholar
Persic, A., Slavic, S., Ravnik, M., and Zagar, T. (2000). TRIGLAV a program package for research reactor calculations. IJS-DP-7862, Jožef Stefan Institute, Ljubljana, Slovenia.Suche in Google Scholar
Poston, D.I. and Trellue, H.R. User’s manual, version 1.00 for Monteburns, version 3.01. United States: N. p., 1998. Web. https://doi.org/10.2172/307942. Los Alamos National Lab. (LANL), USA.Suche in Google Scholar
Sutondo, T. (2014). Analyses of fuel burn-up calculations of kartini reactor based on the new calculation scheme, 17. Center of Accelerator’s Science and Technology (CAST), pp. 91–100.10.17146/gnd.2014.17.2.2801Suche in Google Scholar
Wilson, W., Cowell, S., England, T., Hayes, A., and Moller, P. (2008). A manual for CINDER’90 version 07.4 codes and data. LA-UR-07-8412, Los Alamos National Laboratory, Los Alamos.Suche in Google Scholar
© 2022 Walter de Gruyter GmbH, Berlin/Boston
Artikel in diesem Heft
- Frontmatter
- Effect of engraved concentric circles on pool boiling of water
- Numerical determination of condensation pressure drop of various refrigerants in smooth and micro-fin tubes via ANN method
- Effect of metal layer height on heat transfer inside molten pool
- Transient analysis of MTR research reactor during fast and slow loss of flow accident
- Determination of heat flux leading to the onset of flow instability in MTR reactors
- Experimental study on direct contact condensation of saturated steam at low mass flux in subcooled quiescent water
- Evaluation and integral analysis of ADS and CMT failures during AP1000 SBLOCA with ASYST VER 3 simulation code
- Computational fluid dynamics simulation of material testing reactor spent fuel cooling in wet storage
- The role of advanced nuclear reactors in non-electrical and strategic applications, producing sustainable energy supplies and reducing the greenhouse gasses
- Optimization of PID controller for water level control of the nuclear steam generator using PSO and GA
- Assessment for nuclear security using Analytic Hierarchy Process (AHP) incorporated with Neural Networking Method in nuclear power plants (NPPs)
- Safety assessment and management of spent nuclear fuel for TRIGA mark II research reactor
- Calendar of events
Artikel in diesem Heft
- Frontmatter
- Effect of engraved concentric circles on pool boiling of water
- Numerical determination of condensation pressure drop of various refrigerants in smooth and micro-fin tubes via ANN method
- Effect of metal layer height on heat transfer inside molten pool
- Transient analysis of MTR research reactor during fast and slow loss of flow accident
- Determination of heat flux leading to the onset of flow instability in MTR reactors
- Experimental study on direct contact condensation of saturated steam at low mass flux in subcooled quiescent water
- Evaluation and integral analysis of ADS and CMT failures during AP1000 SBLOCA with ASYST VER 3 simulation code
- Computational fluid dynamics simulation of material testing reactor spent fuel cooling in wet storage
- The role of advanced nuclear reactors in non-electrical and strategic applications, producing sustainable energy supplies and reducing the greenhouse gasses
- Optimization of PID controller for water level control of the nuclear steam generator using PSO and GA
- Assessment for nuclear security using Analytic Hierarchy Process (AHP) incorporated with Neural Networking Method in nuclear power plants (NPPs)
- Safety assessment and management of spent nuclear fuel for TRIGA mark II research reactor
- Calendar of events
