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Oxygen Potential Calculations for Conventional and Cr-Doped UO2 Fuels 2 Based on Solid Solution Thermodynamics

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Oxygen Potential Calculations for Conventional and Cr-Doped UO2 Fuels 2 Based on Solid Solution Thermodynamics 1 Oxygen potential calculations for conventional and Cr-doped UO2 fuels 2 based on solid solution thermodynamics 3 4 5 Enzo Curtia, *, Dmitrii A. Kulika 6 7 8 aLaboratory for Waste Management (LES), Nuclear Energy and Safety Division (NES), 9 Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland) 10 11 12 * Corresponding author. E-Mail address: [email protected] 13 14 15 HIGHLIGHTS 16 Solid solution thermodynamics was successfully applied to model chemical equilibrium in 17 UO2 fuels 18 Cr-doping does not affect significantly the oxygen potential of UO2 fuels 19 Oxygen potentials are sensitive to thermodynamic properties of Mo in -particles 20 Fuel-cladding interaction may affect oxygen potential 21 Oxygen potentials increase with burnup and converge at ≥ 40 GWd/tiHM to values 5-30 22 kJ/mol higher than the Mo/MoO2 buffer 23 24 Keywords 25 Oxygen potential, solid solution, thermodynamics, doped fuel 26 27 28 Figures required in color: 29 Figs. 2, 3, 6, 7, 9 30 This document is the accepted manuscript version of the following article: Curti, E., & Kulik, D. A. (2020). Oxygen potential calculations for conventional and Cr- 31 doped UO2 fuels based on solid solution thermodynamics. Journal of Nuclear Materials, 534, 152140 (18 pp.). https://doi.org/10.1016/j.jnucmat.2020.152140 1 This manuscript version is made available under the CC-BY-NC-ND 4.0 license http://creativecommons.org/licenses/by-nc-nd/4.0/ 32 Abstract 33 34 The effect of Cr-doping on the oxygen potential of UO2 fuels was studied 35 through a comprehensive thermodynamic treatment, relying on the formulation 36 of sublattice solid solutions for the uranium dioxide and for a five-component 37 noble metal phase (, Pd, Ru, Rh, Tc). The UO2 phase includes dissolution of 38 Pu, minor actinides (Am, Cm, Np), soluble fission products (La, Nd, Ce) and 39 trivalent chromium. As far as possible, interaction parameters quantifying non- 40 ideality were calibrated against published experimental data and models. In 41 addition, a (Sr,Ba)ZrO3 ideal solid solution and a number of stoichiometric 42 solids were considered in the equilibrium calculations, as well as ideal gas. The 43 GEM-Selektor code was used for the calculations, with the in-house Heracles 44 database and applying the built-in Berman’s formalism. 45 Ellingham diagrams were calculated for fuel inventories at burnups ranging o o 46 from 1 to 60 GWd/tiHM, over a temperature range (400 C – 1400 C) covering 47 in-pile conditions in light water reactors. The results indicate that Cr-doping will 48 not affect significantly the oxygen potential, except at very low burnup. The 49 calculated oxygen potentials appear to be sensitive to molybdenum dilution in 50 thenoble metal phase,and to a minor extent to its non-ideality. On the contrary, 51 the combined effect of Pu, minor actinides and lanthanides dissolution in UO2 is 52 small. From 20 GWd/tiHM upwards, the calculated oxygen potentials enter the 53 field of experimentally determined values. Above 40 GWd/tiHM, they converge 54 to limiting values about 5-30 kJ/mol higher than the oxygen potentials buffered 55 by coexisting pure metallic Mo and pure MoO2. Further calculations indicate 56 that internal oxidation of the cladding (zirconium alloy) could considerably 57 lower the oxygen potential. 2 58 1. Introduction 59 1.1 Background and objectives of this study 60 In recent years, the use of UO2 fuels doped with small amounts of chromia has become 61 customary in commercial Light Water Reactors (LWR). The dopant induces a coarsening of 62 the UO2 grain size up to about 60 m and a higher densification during the fuel fabrication 63 [1]. The larger grain size reduces Fission Gas Release (FGR), allowing reactors to be operated 64 at higher power than with conventional UO2 fuel and thus for a more efficient energy 65 production. More importantly, the concentrations of volatile fission products in the fuel- 66 sheath gap (such as iodine and cesium) and the yield strength of the fuel are reduced, 67 minimizing the risk of cladding failures through stress corrosion cracking [2, 3]. 68 A collaborative scientific project is currently ongoing in the framework of the Horizon 69 2020 EU program under the acronym DisCo (Modern spent fuel Dissolution and chemistry in 70 failed Container conditions) mainly focusing on spent fuel degradation in a geological 71 repository for radioactive waste [4]. The present contribution is embedded in DisCo ‘s 72 modeling package. The major objective is to compare oxygen potentials of Cr-doped and non- 73 doped UO2 fuel during normal reactor operation in commercial LWRs, in order to predict 74 possible consequences on spent fuel dissolution and on radionuclide release that may arise in 75 a geological repository after containment failure. This objective is pursued by developing a 76 comprehensive thermodynamic model for both types of fuels, capable of taking into account 77 the effects of solid solution formation (in UO2 and in minor phases) and of oxidation at the 78 inner surface of zirconium alloy cladding, as functions of temperature and burnup. Because 79 the focus of the study is on regular reactor operation, excluding the case of severe accidents, 80 the temperature range of calculations was limited to 400-1400 oC. From the model, oxygen 81 potentials vs. temperature plots (Ellingham diagrams) are calculated for both types of fuel. 3 82 The comparison of the predicted oxygen potentials will then allow assessing whether or not 83 Cr-doped fuels would behave unfavorably under repository conditions. 84 85 1.2 The role of oxygen potential ̅ 86 For a system at chemical equilibrium, the partial molar Gibbs energy of oxygen, ∆퐺푂2, is 87 defined as the difference between the chemical potential of free molecular oxygen at a given 88 total pressure p and absolute temperature T, and the same quantity at standard state o o ̅ 표 표 표 89 (T =298.15 K, p = 1 bar) [5]. Hence, ∆퐺푂2 ≡ 휇푂2 − 휇푂2 = 푅푇 ln(푓푂2/푝 ) , where 푝 is the 90 standard state pressure (1 bar) and 푓푂2 is the oxygen fugacity. For the purpose of this 91 application, the oxygen fugacity can be equated to the oxygen partial pressure, 푝푂2, based on 92 the approximation 푓푂2 ≅ 푝 푥푂2 ≡ 푝푂2 , where 푝 is the total pressure and 푥푂2is the mole 93 fraction of O2 in the gas. This simplification is permissible since ideal gas at a low total ̅ 94 pressure of 1 bar are assumed in this study. ∆퐺푂2 is improperly termed “oxygen potential” in 95 most of the literature, which generates confusion since this term actually defines 휇푂2. Because ̅ 96 of the widespread use of this term for ∆퐺푂2, however, it is maintained in this publication. 97 The oxygen potential primarily affects the O/M ratio (i.e. the hyper- or 98 hypostoichiometry) of the uranium dioxide phase, which in turn is responsible for phenomena 99 in the fuel that are highly relevant for reactor operation safety and radionuclide release in a 100 geological repository [5, 6]. For instance, high oxygen potentials favor the oxidation of 101 zirconium alloy at the internal cladding surface, thus increasing the risk of mechanical 102 failures. Moreover, high oxygen potentials could lead to hyperstoichiometric O/M ratios in 103 the UO2 phase, reducing the thermal conductivity of the fuel and thus reactor performance. 104 Hyperstoichiometry also increases the diffusivity of fission products, favoring high Fission 105 Gas Release (FGR) and the segregation of volatile and water-soluble fission products such as 106 Cs and I. These phenomena could lead to high Instant Release Fractions (IRF) after 4 107 containment failure and flooding with aqueous solution in a repository. Moreover, a faster 108 UO2 matrix dissolution rate is expected for hyperstoichiometric fuel [7, 8]. Finally, high 109 oxygen potentials may also lead to the oxidation of multivalent radionuclides (e.g. 235U, 239Pu, 110 93Mo, 126Sn, 237Np), making them more soluble in the aqueous environment of a geological 111 repository. The latter aspects are the most relevant in the context of this contribution to the 112 DisCo project, since they may affect the mobility of dose-relevant nuclides in a geological 113 repository setup and ultimately increase the radiological doses calculated in safety 114 assessments. 115 116 1.3 Previous work and model description 117 A few earlier studies have been carried out to model comprehensively LWR fuel behavior 118 with the aim of determining oxygen potentials [9, 10]. In these studies, typical physical 119 phenomena observed in LWR fuels (e.g., formation of fission gas bubbles, segregation of 120 fission products) were modeled in combination with chemical reactions, predicting the 121 formation of minor metallic and oxide phases. However, chemical reactions were modeled on 122 the generalized assumption of simple ideal solid solution formation [10], which is hardly 123 applicable to the UO2 phase. Moreover, the origin of the underlying thermodynamic data was 124 not documented sufficiently. Clearly, there is an urgent need to improve the chemical models 125 in order to predict reliably oxygen potentials in UO2 fuels. 126 In this study, the focus is on the accurate thermodynamic description of the fuel system. 127 The effects of non-ideal mixing in UO2 and noble metal solid solution (-particles) were 128 considered accurately by calibrating interaction parameters against published experimental 129 data or established mixing models (see Section 3). Segregation processes and transport 130 phenomena were not treated explicitly, but they can to some extent be evaluated indirectly by 131 carrying out calculations at different temperatures representative of the thermal gradient 132 through LWR pellets. 5 133 The final model calculates complete thermodynamic equilibrium in a conventional or Cr- 134 doped UO2 fuel rod irradiated in a LWR, using detailed model inventories with a defined 135 average burnup.
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