Medium Enriched Uranium Research Articles

Concepts for Dry Processing of Spent Nuclear Fuel for Recycling to Light-Water Reactors

We have initiated a study of improved methods for implementing dry (AIROX) processing of commercial spent nuclear fuels for recycling back into light-water reactors. In this proliferation-resistant recycle, the spent fuel is converted to a powder, blended with fresh medium-enriched uranium powder, and refabricated into fuel elements. Evaluations of neutronic characteristics show that it will be necessary to remove a substantial portion of the neutronabsorbing fission products in the spent fuel, especially lanthanides and rhodium, in order to achieve efficient utilization of the spent fuel. We have already modeled oxidative vaporization of selected fission products from the powder at 1000 °C. In addition to permanent gases and fission products that are vaporized during pellet sintering, this can remove Tc, Mo, and some Ru and improve the neutronics. A number of approaches are being evaluated for removing lanthanides and rhodium, initially by thermodynamic modeling and review of literature. The lanthanides exist in solid solution with UO 2 so separations methods will require conversion to fine powder; rhodium is present as a metallic inclusion in the epsilon phase. Chlorination of finely powdered oxide at 1100-1200 °C would vaporize substantial portions of Nd, Eu, Gd, and Rh. A fraction of the uranium would also vaporize; if significant, it could be recovered for recycle. Magnetic and electrostatic methods were evaluated for separation of lanthanide from spent fuel. They are not likely to be practical. However, static separation techniques may be applicable for removing rhodium in the fine powder metallic inclusions. These technicalconsiderations provide the basis for a suggested experimental program.

Advanced Neutron Source Enrichment Study

A study has been performed of the impact on performance of using low-enriched uranium (20% {sup 235}U) or medium-enriched uranium (35% {sup 235}U) as an alternative fuel for the Advanced Neutron Source, which was initially designed to use uranium enriched to 93% {sup 235}U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology.

Analysis of Experiment on Temperature Coefficient of Reactivity in Light-Water-Moderated and Heavy-Water-Reflected Cylindrical Core Loaded with Highly-Enriched-Uranium or Medium-Enriched-Uranium Fuel.

An analysis was performed for the temperature coefficient of reactivity measured in the six types of light-water-moderated and heavy-water-reflected cylindrical cores containing highly-enriched-uranium (HEU) or medium-enriched-uranium (MEU) fuel, which was constructed in the Kyoto University Critical Assembly (KUCA). The purpose of the present analysis was to reveal a mechanism why a light-water region existed in the core contributes to a large positive temperature effect on reactivity. Therefore, based on the assessment of the computational method to calculate the temperature coefficient of reactivity in a multi-region type core, studies were carried out to examine each effect of three physical processes (Doppler broadening, thermal expansion and thermal neutron spectral shift) on the temperature coefficient and to separate each contribution of the multi-regions to this physical quantity. The measured temperature coefficients were approximately simulated by the calculations using the SRAC code system. Th...

Experimental Study on Temperature Coefficient of Reactivity in Light Water-Moderated and Heavy-Water-Reflected Cylindrical Core Loaded with Highly-Enriched-Uranium or Medium-Enriched-Uranium Fuel.

By using the Kyoto University Critical Assembly (KUCA), a series of critical experiments was performed to measure the temperature coefficient of reactivity in a light-water-moderated and heavy-water-reflected cylindrical core loaded with highly-enriched-uranium (HEU) or medium- enriched-uranium (MEU) fuel. The measurement was performed for the approximately 20 to 70°C range to examine the effects of the size of light-water region in a heterogeneous multi- region type core, the reduced 235U enrichment, and the existence of boron burnable poison (BP) on this quantity by using six types of core configurations. In all the six types of cores, there were large light-water regions at the center of core and between the outer fuel region and the heavy-water reflector region, and it was found that these light-water regions caused a remarkably positive effect on the temperature coefficient of reactivity. In the present study, the temperature coefficients of the MEU core and the core without BP were more positive tha...

Nonproliferation Issue of the Pebble Bed High-Temperature Reactor

The constraints of nonproliferation of weapons-grade fuel are most favorably observed in the medium enriched uranium (MEU) fuel cycle of the pebble bed high-temperature reactor, using 20% enriched uranium as feed and thorium as breed material. The cycle can be designed so that the uranium enrichment never exceeds the limitation defined for nonsensitive fuel. In the spent fuel, the amount of fissile plutonium is one order of magnitude lower than for the light water reactor and it is strongly denatured by the even-numbered plutonium isotopes. In the once-through option applied in the introductory phase of the reactor, the proliferation restraints of the plutonium are furnished by the choice of the carbon/heavy metal ratio higher than 450 and of the burnup of 100 MWd/kg heavy metal. The Pu/sub FISS/Pu/sub TOTAL/ is achieved as low as 37%, and the admixing of 8% of /sup 238/Pu would complicate its handling by the decay heat rating. In the closed MEU cycle, the /sup 238/U is continuously separated from the cycle by the use of two different types of fuel elements: Thorium and 20% enriched uranium are inserted into the feed elements, and the uranium recovered from the reprocessing is loaded into the burnup elements,more » without thorium. These elements are removed from the cycle without reprocessing. Again the proliferation risk of the fissile plutonium is minimized because of its very low quantity and high denaturization.« less

Effect of Reducing Fuel Enrichment on the Void Reactivity — Part I: Experimental Study

Reactivity of void in the channels between the fuel plates is measured, and the impact of core conversion from using highly enriched uranium (HEU) to medium-enriched uranium (MEU) in the light-water-moderated cylindrical core with heavy water reflector is investigated on this quantity at the Kyoto University Critical Assembly. The void was generated in the flow channels by producing nitrogen gas bubbles through a small needle-like nozzle and the reactivity effect was measured. The void fraction was measured in an outof-pile experiment. The results indicate that the void effect on reactivity is slightly larger (more negative) in the MEU core than in the HEU core. It is also shown that the interference effect of reactivity by bubbling in two adjacent channels simultaneously is within the experimental error.

Effect of Reducing Fuel Enrichment on the Void Reactivity — Part II: Analytical Study

The results of analyses on the void reactivity measurements performed in the Kyoto University Critical Assembly using medium-enriched uranium fuel as well as highly enriched uranium fuel are provided. In consideration of the heterogeneity of a complex core, four-group constants were generated by SRAC, a standard thermal reactor code system for reactor design and analysis at the Japan Atomic Energy Research Institute. The eigenvalue and perturbation calculations were subsequently performed by the 2D-FEM-KUR code, which is a two-dimensional diffusion code based on the finite element method. The calculated eigenvalue keff agreed with the measured value to within 0.5% in the calculated-to-experiment ratio. The void reactivity calculated by perturbation theory approximately reproduced the experimental data including the spatial dependence. The discrepancy between the calculated and measured void reactivity was <0.05 × 10−3 Δ k / k per voided flow channel.

Analysis of Critical Experiments Using Medium-Enriched-Uranium Fuel in Kyoto University Critical Assembly (KUCA)

The critical experiments using medium-enriched-uranium (MEU) fuel in the Kyoto University Critical Assembly (KUCA), a light-water-moderated and heavy-water-reflected cylindrical core, were started in May 1981, as a part of the international Reduced Enrichment for Research and Test Reactors (RERTR) program. The following KUCA critical experiments were analyzed: (1) the criticality measurements for high-enriched-uranium (HEU) and MEU cores and (2) the reactivity effect measurements of boron burnable-poison (BP) for MEU cores. Five-group constants were generated using the EPRI-CELL code, and two-dimensional diffusion calculations were performed using a conventional finite-difference code DIF3D(2D), and a finite-element code 2D-FEM-KUR. Some of the results from the two diffusion codes were compared with each other. Advantage was taken of the finite-element method for the application of the 2D-FEM-KUR code to a detailed analysis of the BP effect measurements. Differences between the results of calculations and...