Annealing behavior of (Pu,Cm)O2 lattice and bulk expansion from self-irradiation damage

Takano, Masahide; Akabori, Mitsuo; Arai, Y.

doi:10.1016/j.jnucmat.2011.02.060

Cited by 11 publications

(5 citation statements)

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“…This treatment under an Ar/H 2 (4%) flow was composed of a half-hour plateau at 1373 K reached with a heating rate of 10 K·min –1 and followed by a cooling at 25 K·min –1 to room temperature. In order to fully recover the self-irradiation defects, the plateau temperature was selected considering the recovery temperature usually reported for actinide oxides. − The chosen value was also a posteriori ensured by the recent work of Prieur et al on thermal recovery of defects of U 0.8 Am 0.2 O 2−δ , as they obtained a defect-free sample after holding a few minutes at this temperature . The moment when the sample is cooled to room temperature is hence considered to be the experiment start time, called t 0 .…”

Section: Methodsmentioning

confidence: 99%

XRD Monitoring of α Self-Irradiation in Uranium–Americium Mixed Oxides

et al. 2013

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The structural evolution under (241)Am self-irradiation of U(1-x)Am(x)O(2±δ) transmutation fuels (with x ≤ 0.5) was studied by X-ray diffraction (XRD). Samples first underwent a preliminary heat treatment performed under a reducing atmosphere (Ar/H2(4%)) aiming to recover the previously accumulated structural defects. Over all measurements (carried out over up to a full year and for integrated doses up to 1.5 × 10(18) α-decay events·g(-1)), only fluorite U(1-x)Am(x)O(2±δ) solid solutions were observed. Within a few days after the end of the heat treatment, each of the five studied samples was slowly oxidized as a consequence of their move to air atmosphere, which is evidenced by XRD by an initial sharp decrease of the unit cell parameter. For the compounds with x ≤ 0.15, this oxidation occurred without any phase transitions, but for U0.6Am0.4O(2±δ) and U0.5Am0.5O(2±δ), this process is accompanied by a transition from a first fluorite solid solution to a second oxidized one, as the latter is thermodynamically stable in ambient conditions. In the meantime and after the oxidation process, (241)Am α self-irradiation caused a structural swelling up to ∼0.8 vol %, independently of the sample composition. The kinetic constants of swelling were also determined by regression of experimental data and are, as expected, dependent on x and thus on the dose rate. The normalization of these kinetic constants by sample α-activity, however, leads to very close swelling rates among the samples. Finally, evolutions of microstrain and crystallite size were also monitored, but for the considered dose rates and cumulated doses, α self-irradiation was found, within the limits of the diffractometer used, to have almost no impact on these characteristics. Microstrain was found to be influenced instead by the americium content in the materials (i.e., by the impurities associated with americium starting material and the increase of cationic charge heterogeneity with increasing americium content).

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Section: Methodsmentioning

confidence: 99%

XRD Monitoring of α Self-Irradiation in Uranium–Americium Mixed Oxides

et al. 2013

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“…Similarly, future investigations should also include helium accumulation and outgassing studies in the fuel, as well the potential for pellet swelling at operational temperatures. The lattice expansion associated with defects and helium could cause a reduction in thermal conductivity at lower temperatures, as noted by studies with minor actinide bearing fuels [57,58]. Similarly, as was noted by others [23], self-irradiation can influence the heat capacity of a material.…”

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confidence: 75%

Thermal Properties and Behaviour of Am-Bearing Fuel in European Space Radioisotope Power Systems

Watkinson

Mesalam

Vigier

et al. 2021

Thermo

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The European Space Agency is funding the research and development of 241Am-bearing oxide-fuelled radioisotope power systems (RPSs) including radioisotope thermoelectric generators (RTGs) and European Large Heat Sources (ELHSs). The RPSs’ requirements include that the fuel’s maximum temperature, Tmax, must remain below its melting temperature. The current prospected fuel is (Am0.80U0.12Np0.06Pu0.02)O1.8. The fuel’s experimental heat capacity, Cp, is determined between 20 K and 1786 K based on direct low temperature heat capacity measurements and high temperature drop calorimetry measurements. The recommended high temperature equation is Cp(T/K) = 55.1189 + 3.46216 × 102 T − 4.58312 × 105 T−2 (valid up to 1786 K). The RTG/ELHS Tmax is estimated as a function of the fuel thermal conductivity, k, and the clad’s inner surface temperature, Ti cl, using a new analytical thermal model. Estimated bounds, based on conduction-only and radiation-only conditions between the fuel and clad, are established. Estimates for k (80–100% T.D.) are made using Cp, and estimates of thermal diffusivity and thermal expansion estimates of americium/uranium oxides. The lowest melting temperature of americium/uranium oxides is assumed. The lowest k estimates are assumed (80% T.D.). The highest estimated Tmax for a ‘standard operating’ RTG is 1120 K. A hypothetical scenario is investigated: an ELHS Ti cl = 1973K-the RPSs’ requirements’ maximum permitted temperature. Fuel melting will not occur.

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“…The characterization of strain relaxation with temperature has been reported for different UO 2 based materials: UO 2 poly and single crystals implanted with α particles [22][23], self-irradiated MOX [13], U 0.8 Am 0.2 O 2-x [14] or (Pu,Cm)O 2 pellets [15]. For a comparison of these results with the data presented in this paper, we propose to focus on the case of polycrystal Z1 (implanted with 60 keV He ions) which is the simplest mechanical system to analyze and for which the most robust conclusions regarding strain relaxation with temperature have been provided here.…”

Section: Strain Relaxation Versus Temperature: Comparison With Litmentioning

confidence: 99%

“…Such studies have included many aspects: He behavior (lattice location [4], diffusion and precipitation [5]- [7] [8]) but also created defect [9] and the induced strains in the UO 2 matrix [10]. Different approaches have been used to produce relevant samples: aging of UO 2 samples doped with short -lived α-emitters [7][8] [11]- [15], infusion [16], or ion implantation [1] [5].…”

Section: Introductionmentioning

confidence: 99%

“…The relaxation of strains has been analyzed in aged PuO 2 [12], MOX [13], U 0.8 Am 0.2 O 2-x [14] and even (Pu,Cm)O 2 [15] showing that temperatures over 1100 °C are required to get a full strain recovery. As pointed out by Prieur et al, the obtained isochronal annealing curves are not only very close to each other but also close to that obtained for He implanted {111} single crystals [14].…”

Section: Introductionmentioning

confidence: 99%

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Strain relaxation in He implanted UO2 polycrystals under thermal treatment: An in situ XRD study

Palancher

Kachnaoui

Martin

et al. 2016

Journal of Nuclear Materials

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Within the frame of the long-term evolution of spent nuclear fuel in dry disposal, the behavior of He in UO 2 polycrystals has to be studied. In this paper, strain relaxation in He implanted samples have been characterized using in-situ X-ray diffraction during thermal annealing. The influence of a wide range of experimental parameters (annealing atmosphere, He ion energy, orientation of the UO 2 grains probed by X-rays) has been evaluated. If each of them contributes to the strain relaxation kinetics in the implanted layer, strain relaxation is not completed for temperatures below 900 °C which is equivalent to what has been found on He implanted UO 2 single crystals, or aged UO 2 pellets doped with α-emitters. This stands for an additional evidence of the interest of ion implantation to simulate self-irradiation in spent nuclear fuel. Moreover, in the case of implantation with 500 keV He ions, we clearly show that strain relaxation and He release are not correlated for temperatures below 750 °C and demonstrate that this applied in-situ approach is well-suited for measuring strains in ion implanted polycrystals.

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Annealing behavior of (Pu,Cm)O2 lattice and bulk expansion from self-irradiation damage

Cited by 11 publications

References 11 publications

XRD Monitoring of α Self-Irradiation in Uranium–Americium Mixed Oxides

XRD Monitoring of α Self-Irradiation in Uranium–Americium Mixed Oxides

Thermal Properties and Behaviour of Am-Bearing Fuel in European Space Radioisotope Power Systems

Strain relaxation in He implanted UO2 polycrystals under thermal treatment: An in situ XRD study

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