Shojiro Matsuura

A new fuel material for once-through weapons plutonium burning

AbstractFor the burning of plutonium derived from nuclear warheads, once-through type oxide fuels have been studied by considering their proliferation resistance and environmental safety as well as their technological backgrounds of fuel fabrication and reactors.From phase relations of ceramic materials and their chemical properties, it seems that a two-phase mixture of a fluorite-type phase and alumina has favorable characteristics as a once-through-type fuel of plutonium burning. It also seems that the fluorite-type phases such as thoria and fully stabilized zirconia are acceptable as host phases of plutonium because of high solid solubility of the actinide elements and fission products, irradiation stability, and chemical stability. The spent fuels finally obtained will become mineral-like waste forms, which could be buried under deep geological formations without further processing.From reactor burnup calculations with the use of the fuels, light water reactors (LWRs) with the larger volume ratio of m...

Non-Destructive Gamma-Ray Spectrometry on Spent Fuels of a Boiling Water Reactor

The spent fuels from the JPDR-I reactor were measured by means of a γ-scanning facility installed in the fuel storage pool. The spatial distributions of the fission products (134Cs and 137Cs) were measured and analyzed in reference to the effects of control rod pattern. The ratios holding between the products of neutron capture and of direct fission (134Cs/137Cs and 154Eu/137Cs) were also examined for its relevance to non-destructive burnup determination. The activity ratios of the fission products can be expressed by a linear function of burnup, provided that corrections are made to account for differences in irradiation history and for spatial variations in the neutron spectrum.

Buckling Correction for Experimental Power Distribution within Fuel Assemblies in a Finite System

For design calculations to determine the local power distribution in a fuel assembly of a power reactor, the neutron flux is usually assumed to be symmetrical at the outer boundary of the assembly. In actuality, experimental data on power distributions are obtained in a finite system where this symmetry does not apply, so that the calculated values cannot be directly compared with observed data. In a zero power critical experiment in particular, the measurement must be performed in a fairly small core assembly so economize the amount of fuel materials to be used for simulation. This introduces the necessity of special considerations in the comparison between design and observed data. The authors propose a method incorporating direct corrections to the experimentally determined. power distributions based on the geometrical buckling of the system. In this method the experimental power P 0 (r) is divided by the neutron flux Q (r) which is determined in the critical state with geometrical buckling in a bare (...

Measurement of Migration Area and Multiplication Factor of UO2-H2O Lattice

The migration area, multiplication factor, and axial as well as radial reflector savings of a light water moderated critical assembly fueled with UO2 (enriched to 2.596 w/o in 235U) have been obtained through water height experiment applying the one group diffusion model. The water height was varied from 1,226 mm to 710 mm, while the core radius was changed from 183.3 mm to 196.8 mm in order to maintain the criticality. The linear relation obtained between the water height h and (∂p/∂h)-1/3 as well as between the radial buckling Br and the axial buckling B 2 γ indicate that the one group diffusion model is a good approximation for this core.

Measurement of the gamma dose rate distribution in a spent fuel assembly with a thermoluminescent detector

Thermoluminescent detectors (TLDs) were applied to measurements of gamma dose rates and their distributions in spent fuel assemblies irradiated in a boiling water reactor, the Japan Power Demonstration Reactor (JPDR). Materials used for the TLD were Mg/sub 2/SiO/sub 4/:Tb and LiF. Feasibility of the technique was confirmed by an on-site measurement in a spent fuel pond of the JPDR. The highest gamma dose rate calibrated to a /sup 60/Co gamma field was 4 x 10/sup 3/ R/h in the measured positions in a spent fuel assembly, with an average burnup of 5639 MWd/MTM and a cooling time of 8 yr. Reliability of the measurement by means of a TLD was tested via duplicate measurements on a fuel assembly. In the fields of nuclear safeguards and fuel management, application of TLDs will be effective in obtaining inside information of a spent fuel assembly nondestructively.

Measurements of Thermal Disadvantage Factors in Light-Water Moderated PuO2-UO2 and UO2 Lattices

The disadvantage factor for thermal neutrons in light-water moderated PuO2-UO2 and UO2 square lattices were obtained from measurements of thermal neutron density distributions in a unit lattice cell, measured with Dy-Al wire detectors. The lattices consisted of 3.4w/o PuO2-UO2 and 2.6w/o UO2 fuel rods, and the water-to-fuel volume ratio within the unit cell was parametrically changed. The PuO2-UO2 and UO2 fuel rods were designed to realize equal fissile atomic number density. The disadvantage factors thus measured were 1.36±0.07, 1. 37±0.08, 1.40±0.06 and 1.38±0.06 in the PuO2-UO2 fuel lattices, and 1.30±0.06, 1.31±0.08, 1.30±0.08 and 1.33±0.06 in the UO2, for water-to-fuel volume ratios, of 1.76, 2.00, 2.38 and 2.95, respectively. This difference in disadvantage factor between PuO2-UO2 and UO2 fuel lattices corresponds to about 8%. Calculated results obtained by multigroup transport code LASER agreed well with the measured ones.