Noriko Nitani

Thermal expansions of NpO2 and some other actinide dioxides

Abstract Thermal expansions of the stoichiometric actinide dioxides (ThO 2 , UO 2 , NpO 2 and PuO 2 ) were investigated between room temperature and 1300 K using a high temperature X-ray diffraction method. The lattice parameter of NpO 2 at high temperatures was expressed as α T (pm) = 542.03 + 4.28 × 10 −3 T + 9.07 × 10 −7 T 2 − 1.36 × 10 −10 T 3 . Based on excellent reproducible data, thermal expansion of NpO 2 was determined. Thermal expansions of the other actinide dioxides showed good agreement with the respective recommended literature values. The linear thermal expansion coefficients ( α ) of actinide dioxides calculated at 1200 K were in inverse relation to their melting points. At room temperature, however, the α value of UO 2 was found to be higher than those of the other actinide dioxides and this result was discussed.

Thermophysical properties of rock-like oxide fuel with spinel-yttria stabilized zirconia system

Abstract Thermal expansion, thermal diffusivity, melting temperature, Vickers hardness and creep rate of the rock-like oxide (ROX) fuel were measured with the MgAl 2 O 4 (spinel)–ZrO 2 (Y,Gd) (YSZ: stabilized zirconia) system and the MgAl 2 O 4 –YSZ–UO 2 system in the temperature range between room temperature and 1800 K in order to evaluate thermophysical properties. Thermal expansion coefficients of MgAl 2 O 4 –YSZ composites increased with increasing YSZ content and the values were well represented by the Turners equation. Addition of UO 2 to MgAl 2 O 4 –YSZ composite resulted in an increase of thermal expansion. Thermal conductivity values of the MgAl 2 O 4 –YSZ composites decreased with increasing YSZ content and agreed with predictions of the Maxwell–Eucken equation. The eutectic temperature of MgAl 2 O 4 –YSZ and MgAl 2 O 4 –YSZ–UO 2 systems was found to be 2200 K. High temperature hardness of the composites was higher than that of YSZ or MgAl 2 O 4 . The hardness of ROX fuel was considerably higher than that of UO 2 . The creep rate of MgAl 2 O 4 –YSZ composite was controlled by the lattice diffusion of YSZ.

Thermal expansion of neptunium-uranium mixed oxides

Thermal expansions of NpyU{1−y}O2 solid solutions were investigated between room temperature and 1273 K by a high temperature X-ray diffraction technique. The lattice parameters of NpyU1−yO2 solid solutions at high temperatures were given in polynomial expressions of temperature. High temperature heat capacities, Cp of NpyU1−yO2 solid solutions were estimated from the thermodynamic relation using the measured thermal expansions and literature data. The estimated errors in the calculated Cp of UO2 were less than ±5%.

Rock-Like Oxide Fuels and Their Burning in LWRs

Research on the plutonium rock-like oxide (ROX) fuels and their once-through burning in light water reactors has been performed to establish an option for utilizing and disposing effectively the excess plutonium. The ROX fuel is a sort of the inert matrix fuels and consists of mineral-like compounds such as yttria stabilized zirconia, spinel and corundum. A particle-dispersed fuel was devised to reduce damage by heavy fission fragments. Some preliminary results on swelling, fractional gas release and microstructure change for five ROX fuels were obtained from the irradiation test and successive post-irradiation examinations. Inherent disadvantages of the Pu-ROX fuel cores could be improved by adding 238U or 232Th as resonant materials, and all improved cores showed a nearly the same characteristics as the conventional UO2 core during transient conditions. The threshold enthalpy of the ROX fuel rod failure was found to be comparable to the fresh UO2 rod by pulse-irradiation tests simulating reactivity initiated accident conditions.

In-pile irradiation of rock-like oxide fuels

Abstract Five kinds of rock-like oxide fuels were prepared and irradiated using 20% enriched U instead of Pu. The microstructure analyses for irradiated fuel pellets were carried out by ceramography and electron probe microanalysis (EPMA). For MgAl 2 O 4 (spinel)-based fuel, decomposition of the spinel and vaporization of MgO (magnesia) was observed. For Al 2 O 3 (corundum)-based fuels, significant appearance changes were not observed under the irradiation condition of sufficiently high fuel temperature. Fission product (FP) distributions were also analyzed by EPMA. The distributions of Nd and Ba were similar with those of U and Zr. Mo and Pd were found as fine inclusions in the yttria-stabilized zirconia (YSZ) phase. A part of Cs and Xe migrated to the periphery of YSZ particles and to pores in YSZ. The FP distributions of spinel-based fuel were almost similar to corundum-based fuel.

SOLID-LIQUID PHASE EQUILIBRIA OF NP(VI) AND OF U(VI) UNDER CONTROLLED CO2 PARTIAL PRESSURES

Solid-liquid phase equilibria of Np(VI) under 80%, 0.99% and 0.03% C0 2 and of U(VI) under 100% and 0.03% C0 2 were investigated in 0.1 M NaC104 solution of pH ranging 2.9—4.9 at 25±0.1°C. Neptunium was oxidized and stabilized in hexavalent state in the presence of ozone under various C0 2 partial pressures. Aqueous phases of Np(VI) and U(VI) were characterized by pH measurement, liquid scintillation counting, absorption spectroscopy and fluorescence spectroscopy. Solid phases were analyzed by X-ray crystallography, UV-Vis-NIR photoacoustic spectroscopy and FT-IR photoacoustic spectroscopy. Np02C03(s) and U02C0,(s) were found as solubility limiting solid phases under 80% and 100% C02 , while NpO,(s) and U03(s) were characterized as equilibrium solid phases under 0.03% and 0.99% C02 , respectively. The solubility products of the carbonates were found to be lg ^ ( N p O î C O j ) = —14.62±0.12 and lg / f v (U0 2 C0 3 ) = -14.10±0.14, and those of the trioxides were l g / f J N p O , ) = -21.73±0.17 and lgtf,„(U03) = —22.15±0.06.

Irradiation effects on yttria-stabilized zirconia irradiated with neon ions

Abstract In situ TEM observation was performed on yttria-stabilized zirconia during 30 keV Ne + ion irradiation at room temperature, 923 and 1473 K, respectively, and annealing was performed after irradiation. The observed results revealed clear difference in morphology of damage evolution depending on irradiation temperature. In the irradiation at room temperature defect clusters and bubbles were formed homogeneously at random, and defect clusters were formed earlier than bubbles. In the irradiation at 923 K bubbles and dislocation loops are formed heterogeneously almost at the same time. Moreover, bubbles existed almost only on the loop planes but were almost invisible outside of loop planes in the early stage of irradiation. In the irradiation at 1473 K only bubbles were formed and they grew remarkably with the increasing ion fluence. In annealing remarkable growth of bubbles were observed when temperature was raised from 1373 to 1473 K.

Irradiation behavior of rock-like oxide fuels

Two irradiation tests were performed on the rock-like oxide (ROX) fuels in order to clarify in-pile irradiation stabilities. In the first test small disk-shape fuel targets were irradiated in the JRR-3 in JAERI. In the second test pellet-type fuels were employed. Irradiation behaviors such as swelling, fractional fission gas release (FGR) and phase change were examined by puncture test, profilometry and ceramography. Swelling and FGR behavior of the pellet-type fuels improved considerably compared with the disk-type fuels. Yttria stabilized zirconia (YSZ) single-phase fuel showed an excellent irradiation behavior, i.e. low FGR (<3%), negligible swelling and no appreciable restructuring. The particle dispersed fuels showed lower swelling and higher FGR than those of mechanically blended fuels. Spinel decomposition and subsequence restructuring in the spinel matrix fuels was observed for the first time in the present investigation. It would be possible to reduce the FGR of the spinel matrix fuels to that of the corundum ones, if the maximum fuels temperature is limited below 1700 K where neither spinel decomposition nor restructuring was observed. Damaged area of spinel matrix due to fission fragment irradiation seemed to be confined to thin layers around the surface of YSZ particles as expected.

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).

Current status of researches on the plutonium rock-like oxide fuel and its burning in light water reactors

Abstract Intention of the ROX-LWR system research is to provide an option for utilization or disposition of surplus plutonium. Researches on inert matrix materials and irradiation performance shows that the most favorable candidate for the ROX fuel is a particle dispersed fuel where small particles consisted of yttria stabilized zirconia, PuO 2 and some additives are homogeneously dispersed in spinel matrix. Reactor safety analyses show that the ROX fueled PWR core has nearly the same performability as the existing UO 2 fueled PWR under both reactivity initiated accidents and loss of coolant accidents.