Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Investigation of the fission gases behavior in meso scale: application in multiscale simulation of nuclear fuel

Document Type : Research Paper

Authors

M. Abbasi 1
M.H. Porhemmat 1
Y. Taghipour Azar 1
J. Karimi-Sabet , 2

1 Department of Multiscale Simulation-Multiphysics and Computational Analysis, Advanced Research Institute of Simulation and Separation, P.O.Box:5931-143995, Tehran-Iran

2 Nuclear Fuel Cycle Research School, Nuclear Science and Technology Research Institute, AEOI, P.O.Box:11365-8486, Tehran-Iran

https://doi.org/10.24200/nst.2022.1464
Abstract
The behavior of fission gases such as xenon and krypton have a significant effect on the performance of nuclear fuels, their active life,  and the length of the reactor operating cycle. Two important phenomena, including the release of fission gases from the fuel rod matrix and the resulting swelling, are currently the main challenges points in studying the performance of nuclear fuels. In this regard; and in order to multiscale modelling of nuclear fuel a meso-scale fuel performance code has been developed to analyze Fission Gas Release (FGR) and swelling under steady state conditions. The obtained results are compared with experimental data for low burn up to a maximum of 6.5 MWd / tUO2, where good agreement is achieved. In fact, this code is a bridge between data from atomic-scale calculations and macroscopic scales in nuclear fuel performance codes. The developed code, has the capability to be utilized as a standalone code, or to be called as a subroutine in other nuclear fuel performance codes. It is noteworthy that the development of the code for high burnup as well as transient conditions is underway.

Highlights

1. P. Van Uffelen, et al, Analysis of reactor fuel rod behavior, Handbook of Nuclear Engineering (2010).  
 
2. D. Olander, A. Motta, Light Water Reactor Materials, American Nuclear Society Scientific Publications, vol. 1, (2017).   
 
3. P. Van Uffelen, et al, A review of fuel performance modelling, Journal of Nuclear Materials, 516 (2018).
 
4. K. Lassmann, Transuranus: a fuel rod analysis code ready for use, Journal of Nuclear Materials, 188, 295-302 (1992).
 
5. M. Cunningham, et al, Fraptran: a computer code for the transient analysis of oxide fuel rods, Nuregicr-6739, 1 (2001).
 
6. H.S. Aybar, P. Ortego, A review of nuclear fuel performance codes, Progress in Nuclear Energy, 46, 127-41 (2005).
 
7. G. Pastore, et al, Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling, Journal of Nuclear Materials, 456, 398-408 (2015).
 
8. J. Hales, et al., BISON theory manual the equations behind nuclear fuel analysis, Idaho National Lab. (INL), Idaho Falls, ID (United States) (2016).
 
9. D.R. Gaston, et al., Physics-based multiscale coupling for full core nuclear reactor simulation, Annals of Nuclear Energy, 84, 45-54 (2015).
 
10. P. Van Uffelen, et al, Multiscale modelling for the fission gas behaviour in the Transuranus code, Nuclear Engineering and Technology, 43, 477-88 (2011).
 
11. K. Lassmann, C. Walker, J. Van de Laar, Extension of the Transuranus burnup model to heavy water reactor conditions, Journal of Nuclear Materials, 255, 222-233 (1998).
 
12. G. Jomard, et al, Caracas, An industrial model for the description of fission gas behavior in LWR-UO_2 fuel, (2014).
 
13. B. Baurens, et al, 3D thermo-chemical–mechanical simulation of power ramps with Alcyone fuel code, Journal of Nuclear Materials, 452, 578-94 (2014).
 
14. G. Khvostov, Models for numerical simulation of burst FGR in fuel rods under the conditions of RIA, Nuclear Engineering and Design, 328, 36-57 (2018).
 
15. M. Veshchunov, et al, A new mechanistic code SFPR for modeling of single fuel rod performance under various regimes of LWR operation, Nuclear Engineering and Design, 241, 2822-30 (2011).
 
16. M. Veshchunov, et al., Development of the advanced mechanistic fuel performance and safety code using the multi-scale approach, Nuclear Engineering and Design, 295, 116-126 (2015).
 
17. H.G. Joo, et al, Methods and performance of a three-dimensional whole-core transport code DeCART, (2004).
 
18. J. Cho, et al, DeCART v1. 2 User's Manual, (2007).
 
19. B.S. Collins, et al., MPACT VERA Input User s Manual, Version 2.2. 0, Oak Ridge National Laboratory (ORNL) (2016).
 
20. J.A. Turner, et al., The virtual environment for reactor applications (VERA): design and architecture, Journal of Computational Physics, 326, 544-568 (2016).
 
21. G. Pastore, et al, An effective numerical algorithm for intra-granular fission gas release during non-equilibrium trapping and resolution, Journal of Nuclear Materials, 509, 687-99 (2018).
 
22. D. Pizzocri, et al., A model describing intra-granular fission gas behaviour in oxide fuel for advanced engineering tools, Journal of Nuclear Materials, 502, 323-330 (2018).
 
23. D.R. Olander, Fundamental aspects of nuclear reactor fuel elements: solutions to problems, California Univ., Berkeley (USA). Dept. of Nuclear Engineering (1976).
 
24. J. Turnbull, The distribution of intragranular fission gas bubbles in UO2 during irradiation, Journal of Nuclear Materials, 38, 203-212 (1971).
 
25. F.S. Ham, Theory of diffusion-limited precipitation, Journal of Physics and Chemistry of Solids, 6, 335-351 (1958).
 
26. C. Baker, The fission gas bubble distribution in uranium dioxide from high temperature irradiated SGHWR fuel pins, Journal of Nuclear Materials, 66, 283-291 (1977).
 
27. D. Andersson, et al., Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2±x: Implications for nuclear fuel performance modeling, Journal of Nuclear Materials, 451, 225-242 (2014).
 
28. M. Speight, Vacancy potential and void growth on grain boundaries, (1975).
 
29. R. Singh, Isothermal grain-growth kinetics in sintered UO2 pellets, Journal of Nuclear Materials, 64, 174-178 (1977).

Keywords

Mutiscale simulation
UO2 fuel
Fission gas release
Fission gas swelling
Steady state
1. P. Van Uffelen, et al, Analysis of reactor fuel rod behavior, Handbook of Nuclear Engineering (2010).  
 
2. D. Olander, A. Motta, Light Water Reactor Materials, American Nuclear Society Scientific Publications, vol. 1, (2017).   
 
3. P. Van Uffelen, et al, A review of fuel performance modelling, Journal of Nuclear Materials, 516 (2018).
 
4. K. Lassmann, Transuranus: a fuel rod analysis code ready for use, Journal of Nuclear Materials, 188, 295-302 (1992).
 
5. M. Cunningham, et al, Fraptran: a computer code for the transient analysis of oxide fuel rods, Nuregicr-6739, 1 (2001).
 
6. H.S. Aybar, P. Ortego, A review of nuclear fuel performance codes, Progress in Nuclear Energy, 46, 127-41 (2005).
 
7. G. Pastore, et al, Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling, Journal of Nuclear Materials, 456, 398-408 (2015).
 
8. J. Hales, et al., BISON theory manual the equations behind nuclear fuel analysis, Idaho National Lab. (INL), Idaho Falls, ID (United States) (2016).
 
9. D.R. Gaston, et al., Physics-based multiscale coupling for full core nuclear reactor simulation, Annals of Nuclear Energy, 84, 45-54 (2015).
 
10. P. Van Uffelen, et al, Multiscale modelling for the fission gas behaviour in the Transuranus code, Nuclear Engineering and Technology, 43, 477-88 (2011).
 
11. K. Lassmann, C. Walker, J. Van de Laar, Extension of the Transuranus burnup model to heavy water reactor conditions, Journal of Nuclear Materials, 255, 222-233 (1998).
 
12. G. Jomard, et al, Caracas, An industrial model for the description of fission gas behavior in LWR-UO_2 fuel, (2014).
 
13. B. Baurens, et al, 3D thermo-chemical–mechanical simulation of power ramps with Alcyone fuel code, Journal of Nuclear Materials, 452, 578-94 (2014).
 
14. G. Khvostov, Models for numerical simulation of burst FGR in fuel rods under the conditions of RIA, Nuclear Engineering and Design, 328, 36-57 (2018).
 
15. M. Veshchunov, et al, A new mechanistic code SFPR for modeling of single fuel rod performance under various regimes of LWR operation, Nuclear Engineering and Design, 241, 2822-30 (2011).
 
16. M. Veshchunov, et al., Development of the advanced mechanistic fuel performance and safety code using the multi-scale approach, Nuclear Engineering and Design, 295, 116-126 (2015).
 
17. H.G. Joo, et al, Methods and performance of a three-dimensional whole-core transport code DeCART, (2004).
 
18. J. Cho, et al, DeCART v1. 2 User's Manual, (2007).
 
19. B.S. Collins, et al., MPACT VERA Input User s Manual, Version 2.2. 0, Oak Ridge National Laboratory (ORNL) (2016).
 
20. J.A. Turner, et al., The virtual environment for reactor applications (VERA): design and architecture, Journal of Computational Physics, 326, 544-568 (2016).
 
21. G. Pastore, et al, An effective numerical algorithm for intra-granular fission gas release during non-equilibrium trapping and resolution, Journal of Nuclear Materials, 509, 687-99 (2018).
 
22. D. Pizzocri, et al., A model describing intra-granular fission gas behaviour in oxide fuel for advanced engineering tools, Journal of Nuclear Materials, 502, 323-330 (2018).
 
23. D.R. Olander, Fundamental aspects of nuclear reactor fuel elements: solutions to problems, California Univ., Berkeley (USA). Dept. of Nuclear Engineering (1976).
 
24. J. Turnbull, The distribution of intragranular fission gas bubbles in UO2 during irradiation, Journal of Nuclear Materials, 38, 203-212 (1971).
 
25. F.S. Ham, Theory of diffusion-limited precipitation, Journal of Physics and Chemistry of Solids, 6, 335-351 (1958).
 
26. C. Baker, The fission gas bubble distribution in uranium dioxide from high temperature irradiated SGHWR fuel pins, Journal of Nuclear Materials, 66, 283-291 (1977).
 
27. D. Andersson, et al., Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2±x: Implications for nuclear fuel performance modeling, Journal of Nuclear Materials, 451, 225-242 (2014).
 
28. M. Speight, Vacancy potential and void growth on grain boundaries, (1975).
 
29. R. Singh, Isothermal grain-growth kinetics in sintered UO2 pellets, Journal of Nuclear Materials, 64, 174-178 (1977).

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