Enhancing FRAPCON fuel performance code for physical phenomena at high temperature and high burnup
Source abstract
Abstract
Abstract Advanced LWR fuel designs aim for higher burnup and heating rates. Operating at higher fuel temperatures and burnups involve several physical phenomena typically ignored in LWR fuel performance codes. To accurately analyze the fuel rod behavior at such conditions, the modification of existing LWR fuel performance code is necessary. This paper describes the enhancement of FRAPCON, a fuel performance code used by US NRC. In this study, mechanistic models describing porosity, cesium and hydrogen migration and precipitation kinetics were added into FRAPCON-3.5, referred to as FRAPCON 3.5 Enhanced Performance (EP). A survey of MOX fuel thermal conductivity models vs. experimental data was performed and multiplying factors to account for plutonium weight fraction were derived. Overall, the comparison of these revised models showed reasonable agreement with in-pile experiments.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Nuclear Engineering Thermal-Hydraulics
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References (93)
Show source-provided references
1. Geelhood · 2014 · vol. 1
2. Application of the resource-renewable boiling water reactor for TRU management and long-term energy supply
3. The high burn-up structure in nuclear fuel10.1016/S1369-7021(10)70221-2 ↗
4. Caruso · 2007
5. Andrews · 2015
6. Sukjai · 2015
7. Characteristics of composite silicon carbide fuel cladding after irradiation under simulated PWR conditions10.13182/NT12-86 ↗
8. Sukjai · 2014
9. Geelhood · 2014 · vol. 2
10. Thermal conductivity of hypostoichiometric low Pu content (U,Pu)O2-x mixed oxide10.1016/S0022-3115(99)00205-6 ↗
11. A pragmatic approach to modelling thermal conductivity of irradiated UO2 fuel: review and recommendations10.1016/S0022-3115(96)00404-7 ↗
12. Thermal conductivity of uranium–plutonium oxide fuel for fast reactors10.1016/S0022-3115(00)00407-4 ↗
13. Thermal conductivity measurements of high burnup UO2 pellet and a benchmark calculation of fuel center temperature
14. Fuel thermal conductivity: a review of the modeling available for UO2, (U,Gd)O2 and MOX fuel
15. Thermal performance of high burnup fuel – in-pile temperature data and analysis
16. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet10.1016/j.jnucmat.2011.04.020 ↗
17. The effect of plutonium content on the thermal conductivity of (U, Pu)O2 solid solutions10.1016/0022-3115(71)90040-7 ↗
18. Thermal conductivity of near-stoichiometric (U, Pu, Nd)O2 and (U, Pu, Eu)O2 solid solutions10.1016/0022-3115(83)90114-9 ↗
19. Thermal conductivities of (U, Pu, Am)O2 solid solutions10.1016/j.jallcom.2006.12.159 ↗
20. Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor10.1016/j.jnucmat.2008.09.041 ↗
21. Melting temperature and thermal conductivity of irradiated mixed oxide fuel10.1016/0022-3115(93)90203-B ↗
22. Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35MWd/kgHM10.1016/j.jnucmat.2010.03.006 ↗
23. Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up10.1016/j.jnucmat.2011.02.042 ↗
24. Thermal conductivity of (U, Pu)O2−x mixed oxide fuel10.1016/0022-3115(92)90470-6 ↗
25. Oxygen potential and thermal conductivity of (U,Pu) mixed oxides10.1016/j.jnucmat.2005.10.004 ↗
26. A review of the thermophysical properties of MOX and UO2 fuels10.1016/S0022-3115(01)00692-4 ↗
27. Berna · 1997 · vol. 2
28. Lanning · 2005 · vol. 4
29. Development of FRAPCON-EP for high burnup and high temperature fuel pellet behavior modeling
30. Analysis of mixed oixde fuel behavior under reduced moderation boiling water reactor conditions with FRAPCON-EP
31. On the significance of modeling nuclear fuel behavior with the right representation of physical phenomena10.1016/j.nucengdes.2010.10.028 ↗
32. Steady-state fuel behavior modeling of nitride fuels in FRAPCON-EP10.1016/j.jnucmat.2012.04.011 ↗
33. Fundamental aspects of nuclear reactor fuel elements
34. Karahan · 2009
35. Massih · 2006
36. Observations on the release of cesium from UO2 fuel10.1016/S0022-3115(96)00477-1 ↗
37. Development and verifications of fast reactor fuel design code CEPTAR10.13182/NT156-39 ↗
38. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors10.1016/j.jnucmat.2009.11.021 ↗
39. The diffusion coefficients of gaseous and volatile species during the irradiation of uranium dioxide10.1016/0022-3115(82)90419-6 ↗
40. A simple operational gas release and swelling model: II. Grain boundary gas10.1016/0022-3115(80)90224-X ↗
41. A simple operational gas release and swelling model: I. Intragranular gas10.1016/0022-3115(80)90029-X ↗
42. Thermal creep modeling of HT9 steel for fast reactor applications10.1016/j.jnucmat.2010.12.232 ↗
43. Review of HT-9 cladding creep correlations for advanced liquid metal fast reactors
44. Constitutive thermal creep deformation relations for lifetime prediction of a fusion reactor first wall ferritic alloy10.1016/0920-3796(91)90095-8 ↗
45. Performance of HT9 clad metallic fuel at high temperature10.1016/0022-3115(93)90210-P ↗
46. Development of design equations for ferritic alloys in fusion reactors10.1016/0167-899X(85)90007-2 ↗
47. Constitutive design equations for thermal creep deformation of HT-910.1016/0022-3115(84)90576-2 ↗
48. Irradiation effects of HT-9 martensitic steel10.5516/NET.07.2013.706 ↗
49. Stress and temperature dependence of irradiation creep of selected FCC and BCC steels at low swelling
50. Variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400–600 C
51. Sobolev · 2011
52. Todreas · 2012
53. Short-term irradiation behavior of minor actinide doped uranium plutonium mixed oxide fuels irradiated in an experimental fast reactor10.1016/j.jnucmat.2008.12.041 ↗
54. Radial redistribution of actinides in irradiated FR-MOX fuels10.1016/j.jnucmat.2009.01.010 ↗
55. Short-term irradiation behavior of low-density americium-doped uranium–plutonium mixed oxide fuels irradiated in a fast reactor10.1016/j.jnucmat.2010.11.027 ↗
56. Microstructure and elemental distribution of americium-containing uranium plutonium mixed oxide fuel under a short-term irradiation test in a fast reactor10.1016/j.jnucmat.2008.12.028 ↗
57. Restructuring and redistribution of actinides in Am-MOX fuel during the first 24h of irradiation10.1016/j.jnucmat.2013.01.351 ↗
58. Modelling actinide redistribution in mixed oxide fuel for sodium fast reactors10.1016/j.pnucene.2013.10.008 ↗
59. A liquid-metal reactor core demonstration experiment using HT-910.13182/NT93-A17034 ↗
60. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups10.1016/j.jnucmat.2011.02.052 ↗
61. Butterman
62. Fission product behaviour in phenix fuel pins at high burnup10.1016/0022-3115(92)90453-R ↗
63. Observation of a pellet-cladding bonding layer in high power fuel
64. The chemical state of LWR high-power rods under irradiation10.1016/0022-3115(79)90154-5 ↗
65. Characterization of UO2 irradiated in the BR-3 reactor10.1016/S0022-3115(98)00896-4 ↗
66. Chemical states of fission products and actinides in irradiated oxide fuels analyzed by thermodynamic calculation and post-irradiation examination10.15669/pnst.2.5 ↗
67. Fuel-to-cladding gap evolution and its impact on thermal performance of high burnup fast reactor type uranium–plutonium oxide fuel pins10.1016/j.jnucmat.2003.12.015 ↗
68. Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up10.1016/j.jnucmat.2004.01.003 ↗
69. Distributions of volatile fission products in or near the fuel-cladding gap of the FBR MOX fuel pins irradiated to high burn-up10.1016/j.jnucmat.2005.04.054 ↗
70. On the behaviour of cesium and iodine in light water reactor fuel rods10.1016/0022-3115(81)90428-1 ↗
71. Aitken · 1972
72. A model for radial cesium transport in a fuel pellet10.1016/0022-3115(89)90552-7 ↗
73. Evaluation of volatile and gaseous fission product behavior in water reactor fuel under normal and severe core accident conditions10.13182/NT83-A33141 ↗
74. Siefken · 2001 · vol. 4
75. Hagrman · 1981
76. Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment10.1016/j.jnucmat.2015.08.033 ↗
77. Modelling the release behaviour of cesium during severe fuel degradation10.1016/0022-3115(95)00130-1 ↗
78. An analysis method for the fuel rod gap inventory of unstable fission products during steady-state operation10.1016/0022-3115(94)90248-8 ↗
79. Diffusion of cesium in stainless steel under condition of oxygen potential controlled by Mo/MoO2 or NbO2/Nb2O5 oxygen buffer10.1080/18811248.1981.9733244 ↗
80. The diffusion of iodine and caesium in the UO2±x lattice10.1016/S0022-3115(99)00274-3 ↗
81. Release of fission cesium from irradiated UO2 powder10.1080/18811248.1971.9735302 ↗
82. Gauntt · 2010
83. Hales · 2016
84. Cussler · 2009
85. Thermal diffusion of cesium vapor in helium10.1007/BF00829341 ↗
86. Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment10.1016/j.jnucmat.2013.01.339 ↗
87. Simulated UO2 fuel containing CsI by spark plasma sintering10.1016/j.jnucmat.2015.07.030 ↗
88. Mieloszyk · 2015
89. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding10.1016/j.jnucmat.2014.05.013 ↗
90. Hydrogen in Zircaloy-2: its distribution and heat of transport10.1016/0022-3115(60)90004-0 ↗
91. Hydruration du Zircaloy-4 et étude de la distribution de l'hydrogène dans une gaine de combustible REP
92. Multidimensional simulations of hydrides during fuel rod lifecycle10.1016/j.jnucmat.2015.06.037 ↗
93. Lacroix · 2016