Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

Computational Analysis of Burnup and Actinides Accumulation in a Reactor with a Controlled Neutron Spectrum

6/05/2024 2024 - #02 Modelling processes at nuclear facilities

Egorov G.O. Levchenko Y.V. Manturov G.N. Matveev Y.V.

DOI: https://doi.org/10.26583/npe.2024.2.14

UDC: 621.039.51.17; 621.039.5

The paper presents results of calculations of actinide accumulation and transmutation in fuel during operation of a thermal reactor with an oxide uranium-plutonium MOX fuel. The aim is to investigate the uncertainty in calculations of fuel burnup and minor actinide accumulation using different software packages based on a common base of nuclear constants. The research was carried out on the example of burnup calculations of a fuel element model of one of the innovative neutron spectrum controlled reactor with MOX fuel.

The calculations were carried out by using MCNP5 and WIMS-D5 software packages. The obtained results were analyzed and compared with each other. The calculations utilized different nuclear data libraries based on ENDF/B-VII.1 ROSFOND-2020.2. For MCNP5 calculations, the data was presented in an ACE format, whilst for WIMS-D5 calculations the data files were presented in different energy breakdowns in 69 and 172 groups. The calculations of fuel burn-up and minor actinide accumulation by MCNP5 were carried out using two isotopic kinetics modules ORIGEN2 and CINDER90. For these calculations, various nuclear data based on ENDF/B- VII.1 and ROSFOND-2020.2 was used. The performed MCNP5 calculations with ORIGEN2 and CINDER90 modules have shown to be consistent. Compared to MCNP5, the WIMS-D5 program shows that the calculations in 172 groups are more reliable. It’s worth noting that for isotopes 237Np, 238Pu and 243Am there is a sufficiently large uncertainty in determining of their concentrations.


  1. Andrianov A.A., Andrianova O.N., Vu. D.V., Korovin Y.A., Kuptsov I.S., Spiridonova A.A. Specifics of nuclear data testing in the context of the minor actinide transmutation problems. Izvestya Vuzov. Phyzika. 2023, vol. 66, no.7, pp. 13 – 24. DOI: 10.17223/00213411/66/7/2 (in Russian).
  2. Chertovskikh O.I., Belov A.A., Andrianova O.N. Uncertainty Assessment and Sensitivity Analysis in the Problems of Nuclide Kinetic Modeling Using the Sun Module of BPSD. Atomic Energy. 2023, vol. 133, no. 5 – 6, pp. 346 – 349. DOI: https://doi.org/10.1007/s10512-023-01019-x (in Russian).
  3. Davide C., Guglielmo L., Guido M. Comparison Among Monte Carlo Based Burnup Codes Applied to the GFR Demonstrator ALLEGRO. Global Journal of Energy Technology Research Updates. 2018, v. 5, pp. 1 – 10. DOI: https://doi.org/10.15377/2409-5818.2018.05.1
  4. Teplov P., Chibiniaev A., Bobrov E., Alekseev P. The Main Characteristics of the Evolution Project VVER-S with Spectrum Shift Regulation. Proc. of Intern. Conf. PHYSOR-2014, Kyoto, Japan, Sept. 28 – Oct. 3. 2014. Available at: https://inis.iaea.org/search/search.aspx?orig_q=RN:47042783 (accessed Mar.10, 2024).
  5. X-5 Monte Carlo Team, «MCNP – A General Monte Carlo N Particle Transport Code, Version 5», LA- UR-03-1987, 2003. Available at: https://rsicc.ornl.gov/codes/ccc/ccc7/ccc-740.html. Access mode: license in the name of co-author (accessed Mar.10, 2024).
  6. RSICC Computer Code Collection, «WIMS-D5 A Neutronics Code for Standard Lattice Physics Analysis», ORNL, NEA-1507/04 (Nov. 1997). Available at: https://rsicc.ornl.gov/codes/dlc/dlc2/dlc-231.html. Access mode: freely available program (accessed Mar.10, 2024).
  7. Aldama D.L. Documentation for WIMSD-formatted libraries based on ENDF/B-VII.1 evaluated nuclear data files with extended actinide burn-up chains and cross section data up to 2000 K for fuel materials, INDC(NDS)-0674, IAEA, Vienna, 2014. Available at: http://www-nds.iaea.org/publications .
  8. Manturov G.N., Nikolaev M.N., Koscheev V. Nuclear Data for Reactor Neutronics Calculations – ROSFOND Data Library and ABBN-RF Group Data System. Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants. 2021, no. 2, pp.5 – 24. DOI: 10.55176/2414-1038-2021-2-5-24 (in Russian).
  9. Poston D.L., Trellue H.R. User’s Manual, Version 2.0 for MONTEBURNS Version 1.0. LA-UR-99-4999 (Sept. 1999). Available at: https://rsicc.ornl.gov/codes/psr/psr4/psr-455.html. Access mode: license on disk media (accessed Mar.10, 2024).
  10. Wilson W.B., Cowell S.T., England T.R, Hayes A.C., Moller P. et al. A Manual for CINDER’90 Version 07.4 Codes and Data. LA-UR-07-8412 (Dec. 2007, Version 07.4.2 updated March).
  11. RSICC Computer Code Collection, «ORIGEN2.1; Isotope Generation and Depletion Code, Matrix Exponential Method», ORNL, CCC-371 (Dec. 1991). Available at: https://rsicc.ornl.gov/codes/ccc/ccc3/ ccc-371.html. Access mode: license on disk media (accessed Mar.10, 2024).

calculations MOX fuel burn-up actinoids software packages WIMS-D5 MCNP5 ORIGEN2 CINDER90 ENDF/B-VI1 ROSFOND

Link for citing the article: Egorov G.O., Levchenko Y.V., Manturov G.N., Matveev Y.V. Computational Analysis of Burnup and Actinides Accumulation in a Reactor with a Controlled Neutron Spectrum. Izvestiya vuzov. Yadernaya Energetika. 2024, no. 2, pp. 170-184; DOI: https://doi.org/10.26583/npe.2024.2.14 (in Russian).