Tadasumi Muromura

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...

Dissolution study of spent PWR fuel: Dissolution behavior and chemical properties of insoluble residues

Abstract The dissolution behavior of PWR spent nuclear fuels of 7000 to 39000 MWd/t and the chemical properties of fission product insoluble residues obtained by the nitric acid dissolution of the fuels were investigated. UO2 pellets in the irradiated fuel rods (10.7 mm in diameter) sliced to 3 to 5 mm in length were completely dissolved within 2 h in 3M nitric acid at about 100°C, regardless of their burnups. The amount of insoluble residue remaining after dissolution increased linearly with burnup from 7000 to 30000 MWd/t, and above 30000 MWd/t it increased steeply. About 70% of the insoluble residue was composed of fission products such as molybdenum, technetium, ruthenium, rhodium and palladium. The remainder was fine chips of cladding, etc. The relative ratio of these elements in the insoluble residue was different from that in the spent fuel based on calculation. In insoluble residues only hexagonal ruthenium alloy (ϵ-phase) was identified.

Formation of uranium mononitride by the reaction of uranium dioxide with carbon in ammonia and a mixture of hydrogen and nitrogen— I synthesis of high purity UN

Formation of uranium mononitride by the reaction between UO2 and C in an ammonia stream, a mixed 75%H2-25%N2 stream and a mixed 8%H2-92%N2 stream has been studied at 1400–1600°C. For the formation of high purity UN, there exists a minimum mixing ratio (CUO2). It largely depends on the reaction atmosphere and temperature. The mixing ratio in the three atmospheres decreases in the following order: ammonia >75%H2 + 25%N2 > 8%H2 + 92%N2, and it decreases with increasing temperature. The total amount of impurities (carbon + oxygen) in the UN obtained is in the range of 500–1000 ppm. The time necessary for completion of the reaction in ammonia is shorter than in a mixed hydrogennitrogen.

Phase relations in the systems ZrO2Y2O3Nd2O3 and ZrO2Y2O3CeO2

Abstract The phase relations of ZrO 2 Y 2 O 3 Nd 2 O 3 and ZrO 2 Y 2 O 3 CeO 2 systems have been studied at 1100–1600°C. The single region of the fluorite phase was intensively examined using the relation between lattice parameter and composition. In the ZrO 2 Y 2 O 3 Nd 2 O 3 system, 37 mol% Nd 2 O 3 is soluble in Y 2 O 3 -stabilized zirconia (fluorite phase) at 1100°C and 42 mol% Nd 2 O 3 at 1600°C. In the ZrO 2 Y 2 O 3 CeO 2 system, 40 mol% CeO 2 dissolves into the stabilized zirconia at 1600°C.

Metallic phases precipitated in UO2 fuel: I. Phases in simulated fuel

The chemical state of solid fission products (FPs) was studied using simulated spent fuels of 5–30% FIMA, which were made by the heat-treatment at 1273 to 2273 K under various oxygen potentials. The phases formed in the fuel were identified by X-ray diffraction. Under the oxygen potential below −350 kJ/mol O2 at 1673 K, fluorite and perovskite phases were produced with the metallic phases of α and ϵ in the fuel of 10% FIMA. At the potential between −250 kJ/mol O2; and −340 kJ/mol O2 three oxide phases, fluorite + perovskite + scheelite, were formed with three metallic phases, α + ϵ + σ. Above −270 kJ/mol O2 the scheelite and metallic phases (α + ϵ) were precipitated in the fuel matrix. The lattice parameter of the perovskite gradually decreased with oxygen potential from 0.420 nm at −540 kJ/mol O2 to 0.412 nm at −270 kJ/mol O2. There was an abrupt change in lattice parameters of the ϵ phase at the oxygen potential of −300 kJ/mol O2 at 1673 K, which is that of Mo-MoO2 equilibrium. The possible role of Mo on the phase formation was discussed in regard to oxygen potential.

The variation of lattice parameter with hydrogen content of non-stoichiometric plutonium dihydride

Abstract The relation between composition and lattice parameter of non-stoichiometric plutonium dihydride (cub.) has been studied. The samples were prepared by two different methods, namely, thermal decomposition of PuH3.0 and hydrogenation reaction of the lower hydrides. As the H/Pu mole ratio increased from 2·0 to 2·7, the lattice parameter of PuH2+x lineally decreased from 5·360 to 5·340A˚. The influence of the oxidation on the lattice parameter was also given.

Metallic phases precipitated in UO2 fuel

Abstract The formation and the chemical properties of the insoluble residue in simulated spent fuels were studied. Simulated fuels corresponding to burnups of 5 to 30% FIMA were prepared by adding non-radioactive FP elements to uranium in nitrate solution, evaporating the mixed solution to dryness and heating the pellet at 1273 to 2273 K under various oxygen potentials. The insoluble residue was obtained at the dissolution of the simulated fuel with 3M HNO 3 . The chemical composition of the residue was determined by chemical analysis and the phases in the residue were identified by X-ray diffraction. The insoluble residue was composed mainly of Mo and Ru with minor amounts of Pd and Rh. At some conditions Zr and Sn were also included. The amount of residue was less than 1 wt% up to a 10% FIMA burnup, but increased markedly at a higher burnup. The amount of residue decreased at a temperature higher than 1873 K. This was attributed to the fact that the metallic α-phase formed at high temperature was apt to be easily dissolved in HNO 3 . The amount of metallic residue and Mo/Ru ratio decreased with the increase in oxygen potential. The oxygen potential for Mo-MoO 2 equilibrium may be related to this result. Formation of the Zr-molybdate hydrated oxide compound, ZrMo 2 O 7 (OH) 2 (H 2 O) 2 , was observed only in the residue at high burnups and high oxygen potentials. This phase was possibly formed during the dissolution process.

Fluorite type phase in nuclear waste ceramics with high zirconia and alumina contents

In waste ceramics with high zirconia and alumina contents, Y2O3-stabilized zirconia with fluorite structure is the main host phase for actinide and rare earth elements in high-level radioactive waste (HLW). The reactions between the stabilized zirconia and such typical elements in HLW as Cs, Sr, Ce, Nd and U were examined at 1400°C in 4% H2 + 96% He. The solubility of SrO in the stabilized zirconia was considerably low (0.3 wt% SrO), and about 30 wt% Nd2O3 and 15 wt% Ce2O3 were soluble in this phase. A complete solid solution was made between the stabilized zirconia and UO2. When the mixed oxide (Ce, Nd, U)O2−x was allowed to react with the stabilized zirconia, a single phase region of the fluorite structure was found in the composition range of 0–18 wt% mixed oxide, and a two-phase region of the fluorite and pyrochlore in the composition range of 18–58 wt% mixed oxide. At the composition with 60 wt% mixed oxide, only the pyrochlore phase was produced. The phase relations of these oxide systems are also discussed.

Formation of uranium mononitride by the reaction of uranium dioxide with carbon in ammonia and a mixture of hydrogen and nitrogen: II. Reaction rates

Abstract Kinetics of the carbothermic synthesis of UN from UO 2 in an NH 3 stream and a mixed 75% H 2 + 25% N 2 stream were studied in the temperature range of 1400–1600°C by X-ray analysis and weight change measurement of the sample. The weight change was divided into two parts; i.e. weight loss due to carbothermic reduction of UO 2 and weight loss due to removal of carbon by hydrogen. The former followed the first-order rate equation −1 n (1 − α 0 ) = k 0 t , and the latter the rate equation of phase boundary reaction 1 − (1 − α c ) 1/3 = k c t . The apparent activation energy of the former was in the range of 320–380 kJ/mol. The value of the latter in an NH 3 stream was 175–185 kJ/mol, which was smaller than that in a mixed 75% H 2 + 25% N 2 stream (285 kJ/mol). In this method, the rate of the removal of carbon by hydrogen determines that of the formation of high purity UN.

A new idea of excess plutonium once-through burning in light water reactor

Abstract Once-through burning process has been proposed for the disposition of excess plutonium. A new stable fuel material of multi-phases is fabricated based on conventional MOX fuel technologies. After irradiation in LWR, the spent fuels would be geologically stable, and become high level radioactive waste (HLW) without further processing. From the chemical properties and crystal structures, two oxide systems have been proposed: PuO 2 -ThO 2 - Al 2 O 3 -MgO and PuO 2 -ZrO 2 (Y,GD)-Al 2 O 3 -MgO systems. The experimental study has been made to examine the phase relations of the fuel materials and the distribution of fission products. From the burnup calculation study, it was estimated more than 80% of plutonium is transmuted after irradiation and the quality of plutonium becomes very poor. For the zirconia type fuel (PuO 2 ZrO 2 (Y,Gd)-Al 2 O 3 ), the void and Doppler reactivities were calculated to be very small. The two modified fuel systems are therefore proposed to avoid these small reactivities, i.e. the zirconia-thoria fuel system (PuO 2 -(Zr,Th)O 2 -Al 2 O 3 -MgO) and zirconia fuel system with W or Yb additive (PuO 2 -ZrO 2 (Yb)-Al 2 O 3 -MgO-(W).