R.P.C. Schram

Critical evaluation of the thermal properties of Th02 and Th1−yUy02 and a survey of the literature data on Th1−yPuy02

Abstract The thermal properties of Th0 2 and Th 1− y U y 0 2 are critically assessed. These properties include melting point, heat capacity, enthalpy of formation, oxygen potential, thermal conductivity and linear thermal expansion. The literature survey of the thermal properties of Th 1− y PU y O 2 shows a limited amount of data, as a result a critical evaluation is not possible. In addition, the phase diagrams of the binary systems Th0 2 −U0 2 and Th0 2 −Pu0 2 are discussed.

The EFTTRA-T3 irradiation experiment on inert matrix fuels

Abstract In the EFTTRA-T3 irradiation experiment eight uranium bearing inert matrix fuel capsules and eight capsules containing inert matrices have been irradiated. The irradiations were based on the matrices: MgO, MgAl 2 O 4 , Y 3 Al 5 O 12 , Y 2 O 3 and CeO 2 . The heterogeneously dispersed fuel contained fissile inclusions of UO 2 or (U,Y)O x . These capsules were irradiated for 198.87 days in the HFR Petten. The burn-up of the U-bearing capsules was 17–20% FIMA. The present paper describes all aspects of the EFTTRA-T3 irradiation; the fabrication and characterisation of the samples, irradiation and both destructive and non-destructive PIE.

Post Irradiation Examination of Irradiated Americium Oxide and Uranium Dioxide in Magnesium Aluminate Spinel

To study MgAl 2 O 4 spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed 241 AmO x in spinel and 25 wt% micro-dispersed enriched UO 2 in spinel. During irradiation, the initially present 241 Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and 243 Am (4%). The UO 2 spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmO, inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO 2 inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO 2 inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling.

Mechanical behaviour of macro-dispersed inert matrix fuels

Abstract Macro-dispersed inert matrix fuels were irradiated in the high flux reactor in Petten. These fuels consisted of UO 2 inclusions embedded in the inert matrices MgO, MgAl 2 O 4 , Y 3 Al 5 O 12 , CeO 2- x and Y 2 O 3 . The uranium burn-up reached 17.1–19.8% FIMA after an irradiation period of 198.9 days. The sample temperature was about 700–1000 K. Room temperature indentation measurements were performed in the inert matrices before and after irradiation to determine the Vickers hardness and the fracture toughness. The volume swelling of the UO 2 inclusions has been determined. Pellets of UO 2 inclusions embedded in MgO, MgAl 2 O 4 and Y 3 Al 5 O 12 show cracks in the matrix between these inclusions after neutron irradiation. A model is used to describe the fracture behaviour of these inert matrix fuels.

Annealing effects of helium implanted single crystals and polycrystalline magnesium aluminate spinel

Abstract The effects of transmutation produced α-particles and the thermal behaviour of the incorporated helium is simulated by 900 keV 3 He ion implantations at 2 μm depth in (100) surface oriented single crystals and in polycrystalline samples. Thermal Helium Desorption Spectrometry (THDS) and Neutron Depth Profiling (NDP) have been applied to study the release rate of helium and changes in the depth profile, respectively, during isothermal annealing. Release of the helium took place in at least two stages in a wide temperature interval ranging from 600 to 1600 K. For polycrystalline spinel an additional high temperature release stage at T > 1300 K was ascribed to trapping in pores. For the high dose implanted single crystals mainly helium release in bursts was observed due to thermally induced flaking.

Design and fabrication aspects of a plutonium-incineration experiment using inert matrices in a “Once-Through-Then-Out” mode

Abstract In the plutonium incineration experiment, named ‘Once-Through-Then-Out’ (OTTO), that is being prepared by JAERI, PSI and NRG, the use of highly stable inert matrices will be examined. The inert matrices MgAl 2 O 4 spinel and ZrO 2 are insoluble in nitric acid and are considered as good storage media for final disposal. These inert matrices will be used in this experiment, which is representative for an OTTO scenario. A total of 7 Pu-containing targets were prepared for an irradiation in the High Flux Reactor in Petten. The objective of the irradiation is to reach a very high Pu-burnup. The main parameters to be studied are stability under irradiation, swelling, fission gas release and chemical interactions in the fuel. Four targets will be equipped with thermocouples for on-line monitoring of central temperature. Four of the targets contain MgAl 2 O 4 as an inert matrix, 2 targets contain ZrO 2 and one target contains mixed-oxide (MOX) fuel for reference purposes. The fissile plutonium concentration is 0.32–0.44 g cm −3 . Both particle-dispersed fuel and homogeneous dispersions were fabricated in order to test the effect of the size of the fissile inclusions. The design of the experiment and the fabrication of the samples are discussed.

Post irradiation examination of uranium inclusions in inert matrices

The inert matrix materials CeO 2 , MgO, Y 2 O 3 , MgAl 2 O 4 and Y 3 Al 5 O 12 were selected as candidates for inert matrices for the EFTTRA a -T3 neutron irradiation experiment. Most targets contain 20% enriched 235 U fissile inclusions with an average size of roughly 150 μm. The volume fraction of the fissile phase is either 2.5 vol% UO 2 or 19.6 vol% of Y 5.78 UO x in the inert matrices. The samples were irradiated for 198.9 full power days in the High Flux Reactor in Petten. The calculated burn-up is between 17.3 and 19.5% FIMA. The temperature of the cladding was kept at 600 ± 25 K. A dimensional change of at least +5 % is measured after neutron irradiation for Y 3 Al 5 O 12 and MgAl 2 O 4 with macro dispersions of UO 2 ; the other targets with a macro dispersion of UO 2 show a volume change of less than + 1 vol%. The fractional release of the fission gas Xe is more than 40% for the MgAl 2 O 4 and Y 3 Al 5 O 12 matrices with a macro dispersion of UO 2 , the other targets show a fractional release of Xe of less than 15%. Cracks are observed in MgO and MgAl 2 O 4 targets which is possibly related to the stress caused by swelling of the UO 2 inclusions.

Enthalpy increment measurements of cerium mononitride, CeN

Abstract The enthalpy increment of CeN was measured in the temperature range 474–883 K using a drop calorimeter and the X-ray pattern of the sample was recorded at room temperature. A strong increase in the heat capacity was found, which can be related to the linear expansion coefficient of CeN using the Gruneisen relation.