R.A. Verrall

MICROSTRUCTURAL FEATURES OF SIMFUEL : SIMULATED HIGH-BURNUP UO2-BASED NUCLEAR FUEL

Abstract Simulated high-burnup nuclear fuel (SIMFUEL) replicates the chemical state and microstructure of irradiated fuel so that detailed experiments on fission-gas release, thermal conductivity and leaching can be undertaken in the laboratory. Eleven stable elements were added to simulate the compositions of 3 and 6 at% burnup UO2-based fuel. The preparation route featured high-energy grinding and spray drying to achieve homogeneous dispersion of the feed materials on a submicrometre scale. Sintering at 1650°C then provided atom-scale mixing and second-phase development characteristics of high temperature fuels. The gases and volatiles cannot be introduced during fabrication, but can be subsequently added by ion implantation. The microstructure of SIMFUEL was found to be similar to that of irradiated fuel (without bubbles). Spherical metallic Mo-Ru-Pd-Rh precipitates were found uniformly dispersed throughout the matrix. A finely-precipitated perovskite phase was observed decorating the matrix grain boundaries.

Thermal conductivity of hyperstoichiometric SIMFUEL

Abstract The thermal conductivity of hyperstoichiometric SIMFUEL and UO 2+ x was obtained from thermal diffusivity, specific heat and density measurements. The thermal conductivity of UO 2+ x (no fission products present) decreases with increasing O/U ratio; a reduction of 15%, 37% and 56% at 600°C, and 11%, 23% and 33% at 1500°C, was found for O/U ratios of 2.007, 2.035 and 2.084, respectively. For 3 at% SIMFUEL there was little difference in the thermal conductivity of specimens annealed at oxygen potentials ( ΔG O 2 of -540 kJ/mol (corresponding to UO 2.000 ) and -245 kJ/mol (corresponding to UO 2.007 ). For SIMFUEL, annealed in reducing conditions, the fission products lowered the thermal conductivity significantly. However, for high oxygen potentials ( ΔG O 2 ≥ −205 kJ/mol), the thermal conductivities of UO 2+ x and SIMFUEL were found to be approximately equal in the temperature range of 600 to 1500°C. Consequently, excess oxygen is the dominant factor contributing to thermal conductivity degradation at high oxygen potentials.

Summary of experimental results for ceramic breeder materials

Abstract Lithium-containing ceramics were quickly recognized as promising tritium breeding materials for fusion reactor blankets, particularly because of their safety advantages. Relevant material properties were investigated to evaluate further their suitability. An extensive RD overall properties (baseline, thermal, mechanical); compatibility with structures and beryllium; tritium release characteristics; irradiation behavior; activation; reprocessing; waste disposal issues. As a result of this investigation, lithium-containing ceramics are considered to be excellent tritium breeding materials.

Thermal conductivity of SIMFUEL

Abstract Thermal diffusivity and specific heat of SIMFUEL — simulated high-burnup UO 2 fuel with an equivalent burnup of 3 and 8 at% — were measured between 300 and 1800 K, and the data were combined to obtain the thermal conductivities. The thermal conductivity of SIMFUEL provides a model for the intrinsic conductivity of high-burnup fuels (i.e. without gas bubbles). The results on conductivity of 3 and 8 at% burnup SIMFUEL were lower than those measured for UO 2 by 29 and 45% at 300 K and by 6 and 15% at 1773 K. The reduction in thermal conductivity was approximately linear with burnup. The change in thermal conductivity is a possible explanation for the enhanced gas release observed from high-burnup fuel.

Modelling of UO2-based SIMFUEL thermal conductivity The effect of the burnup

Abstract Thermal conductivities of stoichiometric SIMFUEL — simulated high-burnup UO 2 fuel — with equivalent burnups of 1.5, 3 and 8 at%, deduced from thermal diffusivity and specific heat measurements, are modelled as a function of burnup, taking into consideration two key microstructural features of irradiated high-burnup fuel: 1. (i) the precipitated fission-product phases, and 2. (ii) the dissolved fission products in the matrix. The degradation of the UO 2 matrix thermal conductivity due to the dissolved fission products (after corrections for the precipitated-product phases) was analyzed by considering their effect on the phonon heat current. The degree of the scattering of the phonons by dissolved additives in the matrix (scattering parameter) is a function of the phonon frequency, and was determined using phonon heat current theory. The scattering parameter followed the theoretically predicted square root dependence on the temperature and concentration of the dissolved additives. The thermal conductivity of the SIMFUEL matrix, predicted from the model based on the phonon scattering by dissolved fission products taken into account their mass difference, was in good agreement with the measured values. These results indicate that the degradation of fuel thermal conductivity, in the absence of fission-gas bubbles, can be explained by the phonon scattering from the dissolved fission products.

Canadian ceramic breeder technology: recent results

Abstract Pebble bed ceramic breeders have been under development in Canada for over ten years. The goal is to fabricate and characterize these materials for use in engineering test reactors and in subsequent fusion power reactors. The program emphasis is on 1.2 mm diameter Li2ZrO3 and Li2TiO3 pebbles. Practical use of these pebbles requires a mass-production fabrication process, and characterization of the pebble beds with respect to bed behaviour and irradiation effects. This paper summarizes the relevant work within Canada since 1991. The fabrication process presently used is suitable for mass production, and is in the process of being transferred to industry. Thermal cycling tests have been conducted on zirconate and titanate pebbles under both laboratory and “engineering” conditions. Cycling reduces the pebble strength, although there are indications that different fabrication conditions produce more robust pebbles. This is an active area of work. Single-size lithium zirconate pebbles have been well-characterized in terms of the bed thermal conductivity and purge gas pressure drop. Recent results include measurement of thermal conductivity from 100 to 1200°C (and 0–2 bar), and of purge gas pressure drop as a function of porosity. Binary beds have also been studied, using steel or lithium zirconate smaller pebbles. Extensive irradiation testing of the as-fabricated ceramic is a critical factor in their acceptance. Lithium zirconate has been characterized under several European irradiation tests, and 1.2 mm lithium zirconate pebbles have been tested to 5.2% lithium atom burnup and over 250–1000°C in the BEATRIX-II and CRITIC-2 purged-capsule experiments. Tritium release is rapid even at low temperatures, with no effects of burnup seen. The pebble bed temperature has been consistent with model predictions, and stable under irradiation. Post-irradiation anneal tests of lithium titanate show good tritium release. Post-irradiation examination of the BEATRIX-II lithium zirconate pebbles is just beginning. Reference blanket designs have been developed based on breeder-in-tube geometry. Engineering-oriented tests have been carried out on large-volume (41) and long-pin (3 m) geometries, to characterize the behaviour of the pebble beds under more realistic conditions. The results of the work described here, and related tests within the world fusion community, continue to support the use of these ceramic breeder pebbles in fusion reactor blankets.

Specific heat measurements of UO2 and SIMFUEL

Abstract New specific measurements are reported for UO 2 with eleven additives, simulating 8 at.% burnup, at temperatures between 300 and 1673 K. A Netzsch DSC 404 heat-flux differential scanning calorimeter was used. The results confirm our earlier measurements, showing no anomalous increase relative to UO 2 .

EXOTIC-7: Irradiation of ceramic breeder materials to high lithium burnup

Abstract The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li 2 ZrO 3 , LiAlO 2 and Li 8 ZrO 6 and pebbles of Li 4 SiO 4 and Li 2 ZrO 3 , with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li 4 SiO 4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented.

Irradiation of lithium zirconate pebble-bed in BEATRIX-II Phase II

Abstract BEATRIX-II was an in-situ tritium recovery experiment that was designed to characterize the behavior of lithium ceramics irradiated to a high burnup, and to assess their suitability for use in a fusion reactor blanket. This paper describes the results from the vented canister containing 29.47 g of lithium zirconate spheres packed in a bed 13.2 mm OD, 2.3 mm ID and 103 mm long. The enriched lithium spheres (85% 6Li) were irradiated to a burnup of 5.2% (total lithium) in a steep temperature profile −400°C edge, 1100°C center. The sweep gas was He-O.1% H2, with systematic tests using alternate compositions: He-0.01% H2 and pure He (maximum duration 8 days). Tritium recovery decreased slightly at lower H2 concentrations; for example, the buildup of inventory during a 4-day test in pure He was 0.8 Ci, approximately 6.5% of the tritium generated in the lithium zirconate during that period. The steadiness of the bed central temperature and the tritium release rate, together with low moisture release indicate good performance of the zirconate bed.

Performance of a Li2ZrO3 sphere-pac assembly in the CRITIC-II irradiation experiment

Abstract The performance of a Li2ZrO3 sphere-pac assembly is being evaluated in a long-term, temperature-gradient irradiation experiment, CRITIC-II. After 100 full-power days of operation, the steady-state recovery rate remains at ∼ 2 Ci/day (1 Ci = 37 GBq) with an edge temperature of 300°C and the reference sweep gas of He/0.1% H2. The temperature gradient across the bed assembly is 650°C. Tritium release and inventory changes are being monitored as a function of temperature, sweep-gas composition, and irradiation time. Operation at a minimum temperature of 200°C resulted in an inventory build-up of 3.3 Ci after 15.8 FPD, with the tritium recovery rate approaching the steady-state release rate. To date, the Li2ZrO3 sphere-pac assembly has demonstrated good performance as a candidate blanket material.