Model of nuclear fuel pellets densification under irradiation and isothermal conditions: Application to UO2 fuels
Source abstract
Abstract
Abstract The dimensional changes of a nuclear fuel in operation are strongly determined by two opposite effects. One of them is due to contraction of the as-fabricated pores, giving place to densification which is evident during the first stages of irradiation. This effect is counteracted by the swelling phenomenon provoked by the fission products that progressively accumulate in the fuel material. A model to evaluate the changes in fuel pellets porosity due to radiation and thermal effects taking into account the point defects flow to and from the pores is presented. A simplification of the model to assess the progress of porosity in isothermal re-sintering tests is also given. Simulations are compared with experimental data measured on UO2 fuel pellets with a variety of microstructures at different temperatures and radiation conditions, attaining a good agreement.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Radioactive element chemistry and processing
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (41)
Show source-provided references
1. Olander · 1976
2. Thermal and in-reactor densification of UO2: mechanisms and experimental results10.1016/0029-5493(78)90208-X ↗
3. Irradiation-induced densification of UO2 pellet fuel10.1016/0022-3115(76)90013-1 ↗
4. In-pile intragranular densification of oxide fuels10.1016/0022-3115(78)90454-3 ↗
5. In-pile densification of intergranular porosity in oxide fuels
6. A model for predicting evolution of pore size distributions in UO2 fuel10.1016/0022-3115(91)90401-R ↗
7. On the rate of in-reactor UO2 densification10.1016/0022-3115(77)90188-X ↗
8. Model for evolution of crystal defects in UO2 under irradiation up to high burn-ups10.1016/j.jnucmat.2008.09.024 ↗
9. Models for fuel porosity evolution in UO2 under various regimes of reactor operation10.1016/j.nucengdes.2014.02.016 ↗
10. A new characterization approach for studying relationships between microstructure and creep damage mechanisms of uranium dioxide10.1016/j.jnucmat.2016.02.027 ↗
11. The surface energy of UO2 as determined by hertzian indentation10.1016/0022-3115(80)90048-3 ↗
12. Irradiation behaviour of fission-gas bubbles and sintering pores in UO210.1016/0022-3115(69)90175-5 ↗
13. Modelling of as-fabricated porosity in UO2 fuel by MFPR code
14. A new fission-gas release model10.1016/0022-3115(83)90176-9 ↗
15. Representation and evaluation of particle size analysis data10.1002/ppsc.19840010115 ↗
16. Model for evolution of as-fabricated pores in UO2 fuel under irradiation conditions
17. The Dependence of in-reactor UO2 densification on temperature and microstructure10.1016/0022-3115(74)90176-7 ↗
18. In-Reactor UO2 densification10.1016/0022-3115(76)90273-7 ↗
19. Simulation of isothermal gas release10.1016/0022-3115(95)00203-0 ↗
20. A comparison between fission gas release data and FEMAXI-IV code calculations10.1016/0029-5493(87)90055-0 ↗
21. Bouloré · 2001
22. A model for fission gas release and gaseous swelling of the uranium dioxide fuel coupled with the FALCON code10.1016/j.nucengdes.2011.06.020 ↗
23. The theory of sink strengths
24. Interaction of dislocations in UO2 during high burn-up structure formation10.1016/j.jnucmat.2013.09.042 ↗
25. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods10.1016/j.jnucmat.2007.06.020 ↗
26. On the coefficient for bulk recombination of vacancies and interstitials
27. The response of fission gas bubbles to the dynamic behavior of point defects10.1016/0022-3115(79)90223-X ↗
28. A dynamic intragranular fission gas behavior model
29. Investigation of cation self-diffusion mechanisms in UO2±x using molecular dynamics10.1016/j.jnucmat.2013.08.030 ↗
30. Technological problems and the future of research on the basic properties of actinide oxides10.1039/f29878301273 ↗
31. Radiation enhanced diffusion in UO2 and (U, Pu)O210.1080/00337578308224715 ↗
32. Atomic transport properties in UO2 and mixed oxides (U, Pu)O210.1039/f29878301121 ↗
33. Atomic mechanisms of mass transport in ceramic nuclear fuel materials10.1039/ft9908601243 ↗
34. Uranium and oxygen self-diffusion in UO210.1016/0022-3115(69)90164-0 ↗
35. Grain-boundary diffusion in uranium dioxide: the correlation between sintering and creep and a reinterpretation of creep mechanism10.1016/0022-3115(79)90035-7 ↗
36. Thermodynamics vol. II
37. Lattice and grain-boundary diffusion of uranium in UO210.1016/0022-3115(66)90004-3 ↗
38. Calculation of diffusion penetration curves for surface and grain boundary diffusion10.1063/1.1699825 ↗
39. Brite · 1975
40. Derivation of analytical expressions for the network dislocation density, change in lattice parameter, and for the recrystallized grain size in nuclear fuels10.1016/j.jnucmat.2005.10.007 ↗
41. Lide · 2003 · 3485