Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Spent fuel characterization analysis using various nuclear data libraries

  • Received : 2022.02.01
  • Accepted : 2022.04.14
  • Published : 2022.09.25
Download PDF
⟨ Previous Next ⟩

Abstract

Experience shows that the solution to waste management in any national programme is lengthy and burdened with uncertainties. There are several uncertainties that contribute to the costs associated with spent fuel management. In this work, we have analysed the impact of the current nuclear data on the isotopic composition of the spent fuel and consequently their influence on the main spent fuel observables such as decay heat, activity, neutron multiplication factor, and neutron and photon source terms. Nuclear libraries based on the most general nuclear data ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 are considered. A typical NPP Krško fuel assembly is analysed using the Monte Carlo code Serpent 2. The analysis considers burnup of up to 60 GWd/tU and cooling times of up to 100 years. The comparison of results showed significant differences, which should be taken into account when selecting the library and evaluating the uncertainty in determining the characteristics of the spent fuel.

Keywords

Acknowledgement

The authors acknowledge that the work presented here was financially supported by the Euratom Research and Training Program under Grant Agreement No. 847593 and the IAEA coordinated research project T13018 - Spent Fuel Characterization.

References

  1. https://sloveniatimes.com/nek-building-dry-cask-storage/.
  2. E. Vlassopoulos, B. Volmert, A. Pautz, Logistics optimization code for spent fuel assembly loading into final disposal canisters, Nucl. Eng. Des. 325 (2017) 246. https://doi.org/10.1016/j.nucengdes.2017.04.036
  3. V. Solans, D. Rochman, H. Ferroukhi, A. Vasiliev, A. Pautz, Loading optimization for Swiss used nuclear fuel assemblies into final disposal canisters, Nucl. Eng. Des. 370 (2020).
  4. J. Leppanen, et al., The Serpent Monte Carlo code: status, development and applications in 2013, Ann. Nucl. Energy 82 (2015) 142-150, 2015. https://doi.org/10.1016/j.anucene.2014.08.024
  5. Kang-Seog Kim, William A. Wieselquist, Neutronic characteristics of ENDF/BVIII.0 compared to ENDF/B-VII.1 for light-water reactor analysis, J. Nucl. Eng. 2 (2021) 318-335. https://doi.org/10.3390/jne2040026
  6. D. Rochman, et al., Nuclear data uncertainties for typical LWR fuel assemblies and a simple reactor core, Nucl. Data Sheets 139 (2017) 1-76. https://doi.org/10.1016/j.nds.2017.01.001
  7. Matthias Frankl, Mathieu Hursin, Dimitri Rochman, Alexander Vasiliev, Hakim Ferroukhi, Nuclear data uncertainty quantification in criticality safety evaluations for spent nuclear fuel geological disposal, Appl. Sci. 11 (14) (2021).
  8. Bamidele Ebiwonjumi, Chidong Kong, Peng Zhang, Alexey Cherezov, Deokjung Lee, Uncertainty quantification of PWR spent fuel due to nuclear data and modeling parameters, Nucl. Eng. Technol. 53 (3) (March 2021) 715-731. https://doi.org/10.1016/j.net.2020.07.012
  9. M. Kromar, B. Kurincic, Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 libraries for the nuclear design calculations of the NPP Krsko with the CORD-2 system, ASME J of Nuclear Rad Sci (April 27, 2021).
  10. M. Pusa, J. Leppanen, An ef ficient implementation of the Chebyshev rational approximation method (CRAM) for solving the burnup equations, in: Proc. PHYSOR 2012, Knoxville, TN, Apr. 15-20, 2012, 2012.
  11. A. Isotalo, P. Aarnio, Substep methods for burnup calculations with Bateman solution, Ann. Nucl. Energy 38 (2011c) 2509-2514, 2011. https://doi.org/10.1016/j.anucene.2011.07.012
  12. M.B. Chadwick, et al., ENDF/B-VII.0: next generation evaluated nuclear data library for nuclear science and technology, Nucl. Data Sheets 107 (2006) 2931. https://doi.org/10.1016/j.nds.2006.11.001
  13. M.B. Chadwick, et al., ENDF/B-VII.1: nuclear data for science and technology: cross sections, covariances, fission product yields and decay data, Nucl. Data Sheets 112 (2011) 2887. https://doi.org/10.1016/j.nds.2011.11.002
  14. D.A. Brown, et al., ENDF/B-VIII.0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data, Nucl. Data Sheets 148 (2018) 1. https://doi.org/10.1016/j.nds.2018.02.001
  15. A.J.M. Plompen, et al., The joint evaluated fission and fusion nuclear data library, JEFF-3.3, Eur. Phys. J. A 56 (2020) 181. https://doi.org/10.1140/epja/s10050-020-00141-9
  16. T. Kaltiaisenaho, Photon transport physics in Serpent 2 Monte Carlo code, Comput. Phys. Commun. 252 (2020) 107143. https://doi.org/10.1016/j.cpc.2020.107143
  17. T. Kaltiaisenaho, Photonuclear reactions in serpent 2 Monte Carlo code, in: Proc. Int. Conf. On Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Portland, USA, August 25-29, American Nuclear Society, 2019, pp. 181-190.
  18. G. Allen, Croff, ORIGEN2: a versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials, Nucl. Technol. 62 (3) (1983) 335-352. https://doi.org/10.13182/NT83-1
  19. H.G. Hughes, Improvements in Electron-Photon-Relaxation Data for MCNP6, vols. 16-20840, RPSD-2016, Paris, France, LA-UR, 2016.