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Four pins of U-Pu mixed carbide fuel, which is one of potential advanced fuels for FBRs, were irradiated in JRR-2 up to burn-ups of 1.5% FIMA without any pin failure. In spite of irradiations at relatively low linear heat powers of 42 and 64kW/m, the change in the sizes of pores and grain size was observed. Pellet density and carbon content appreciably influenced to irradiation fuel performance, such as restructuring, cracking and FP gas release. It was revealed in the γ-spectrometry that Cs behaved similarly to FP gases and condensed at the plenum. A slight carburization was observed at the inside of 316 stainless steel cladding.

A simple and accurate method for the determination of nitrogen in uranium- and plutonium-bearing materials was developed. The loss of nitrogen by oxidation of the sample before analysis was prevented by pulverizing, weighing and packing the sample into a tin capsule in a glove-box with a high-purity argon atmosphere. Nitrogen was determined by oxidizing the nitrides in the tin capsule in circulating oxygen at 850 °C and analysing the combustion gases by gas chromatography with thermal conductivity detection. The relative standard deviation was about 0.7% and the time required to analyse one sample was about 10 min for successive analyses. Skilled techniques for glove-box work are not necessary. The method is applicable not only to the analysis of research samples but also to the quality control of nitride fuel production lines.

The thermal conductivities of (U, Pu)C pellets containing MoC up to 10 mol% were determined from 680 to 1600 K by the laser flash method. The thermal conductivities decreased with MoC content within a single phase region of (U, Pu, Mo)C solid solution. On the other hand, they kept at an almost constant value above the solubility limit which was about 3.5 mol% of MoC in (U, Pu)C. The contribution of electrons to the thermal conductivity was analyzed by the results of the electrical resistivity measurements. The thermal conductivities of the quasi-ternary compounds, (U, Pu)MoC2 and (U, Pu)MoC1.7, were also determined.

The effect of solid fission products on the thermal conductivity of ovide fuels has been studied systematically by simulating them with their stable isotopes. (U, R)O2.00 (R ev Nd, Sm, Eu, Gd, Y; RO1.5 = 0–15 mol.%), (Pu, R')O2.00 (R' ev Nd, Y; R'O1.5 ev 0–10 mol.%) and (M, R)O2.00 (M = U0.8 + Pu0.2 R ev Nd, Eu; RO1.5 ev 0–10 mol.%) were prepared and thermal conductivities were determined in the temperature range of 700–1900 K from thermal diffusivities measured by a laser flash method. The decreases in the thermal conductivities of solid solutions by adding rare earth elements to the ovide fuels are governed mainly by the lattice defect thermal resistivities. These can be reasonably evplained by the lattice defect model for dielectric solids considering U4+, U5+, Pu4+, Pu5+, R3+ or O2− ions as the phonon scattering centres.

The thermal conductivities of near-stoichiometric (U, Ce)C and (U, Pu, Ce)C solid solutions containing CeC up to 10 mol% were determined in the temperature range from 740 to 1600 K by the laser flash method. The thermal conductivity decreased with the cerium content in the solid solutions. The electrical resistivities were also measured for the purpose of analyzing the heat conduction mechanism. It was found that the decrease of electronic heat conduction caused by the addition of cerium resulted in decreasing the thermal conductivities of (U, Ce)C and (U, Pu, Ce)C compared with UC and (U, Pu)C.

The release of the occluded gases in uranium-plutonium mixed carbide and uranium carbide pellets for irradiation tests were measured successfully using a multipurpose high temperature apparatus. A new type of gas line filter was set up in the gas exhausting line, where the filter could be exchanged easily without plutonium contamination and degassed by heating. The release of the occluded gases in the carbide fuels was measured at 1,473, 1,673 and 1,873 K. At 1,473 K, most of the gases evolved was H2 gas and above this temperature, the oxygen contained in the fuel was released continuously in the form of CO gas. It was shown that the H2 gas was generated by the reaction of the fuel with moisture adsorbed at the fuel surface during storage and the CO gas was released in proportion to its equilibrium pressure of the oxycarbide solid solution. The release of the gases from the fuels scarcely affects the increase of the inner gas pressure of fuel pins since the amount of H2 gas is small, about 100 μl·atm/g-fu...

Calcul des conductivites thermiques de ces composes de l'uranium a partir de leurs diffusivites thermiques mesurees entre 750 et 1500°K

The electrical conductivity of near-stoichiometric (U, Gd)O2 solid solutions containing gadolinia up to 14 mol% was measured by four probes dc technique. The addition of gadolinia to urania enhanced the p-type electrical conductivity in the temperature range from room temperature to 1000 K. The electrical conductivity was able to be interpreted in terms of hopping of localized holes between U5+ and U4+ ions in the solid solutions. The experimental results presented in this study were consistent with the behavior predicted by the adiabatic small polaron theory.

Uranium-plutonium mixed nitride, (U, Pu)N, was fabricated by the carbothermic reduction in a N 2 -8% H 2 mixed gas stream from UO 2 and PuO 2 . The influence of the sintering conditions, such as milling time, compacting pressure, sintering temperature, sintering atmosphere and pore former content, on the characteristics of (U, Pn)N pellets was systematically investigated. High density (~ 95% TD) pellets were obtained by sintering at 2023 K in an Ar-8% H 2 or Ar stream without loss of plutonium using the highly activated powder. On the other hand, low density (~ 85% TD) pellets with excellent thermal stability were fabricated by sintering with pore former particles. The difference in the microstructure or composition was discussed in connection with the sintering conditions. Certain increase in the lattice parameter of (U, Pu)N was observed during sintering stage.

The thermal conductivities of stoichiometric (Pu, R)O 2 solid solutions containing RO 1.5 (R = Nd and Y) up to 10 mol% were determined by the laser flash method in the temperature range 700–1400 K. The thermal conductivities for all solid solutions satisfied the phonon conduction equation K = ( A + BT ) −1 within ± 6%. The lattice defect thermal resistivities (= A ) increased gradually with the neodymium or yttrium content, while the intrinsic lattice thermal resistivities (= BT ) were nearly equal to that measured for PuO 2 . The measured thermal conductivities were consistent with the results of the lattice defect model calculation in which Pu 4+ , Pu 5+ and R 3+ ions were considered as phonon scattering centers. The lattice strain parameters ϵ = 85 and 93 were obtained for (Pu, Nd)O 2 and (Pu, Y)O 2 solid solutions, respectively.

A new method for the determination and control of the carbon potential of a sample material using CH 4 /H 2 gas equilibration was developed. The carbon potential of the sample could be controlled in a wide range with an accuracy of about ±0.1 kcal/mol by adopting several kinds of reference mixtures, in which the temperature of the reference material, metal/metal-carbide mixture set, was changed. The carbon potential in austenitic 316 stainless steel at 1000°C was measured successfully by this method.

The evaporation mechanism and the rate-controlling step were clarified for polycrystalline Th 0.90 U 0.10 O 2.05 and Th 0.75 U 0.25 O 2.13 solid solutions. The evaporation annealing was performed under a flowing air atmosphere at 1650°C for 32 h. The solid solutions showed incongruent evaporation in which uranium oxide preferentially vaporized. The rate of the vaporization was found to be controlled by the cation diffusion in the solid phase and enhanced by the presence of grain boundaries. In the case of the UO 2 -rich solid solution many cracks and pores were produced beneath the surface due to the enhancement.

The thermal conductivities of (Pu1−xRx)O2−y solid solutions (R = Nd and Y) containing RO1.5 up to 10 mol% were determined in the temperature range 700–1450 K from thermal diffusivities measured by the laser flash method. The thermal conductivities satisfied the phonon conduction equation K = (A + BT)−1 within ± 7%. The values of A, corresponding to the lattice defect thermal resistivity, increased linearly with the neodymium or yttrium content, while those of B were nearly constant. The increasing rate of A for (Pu, Nd)O2−y solid solutions was slightly larger than that for (Pu, Y)O2−y. These increases were reasonably explained by the lattice defect model in wich Pu4+, R3+, O2− ions, and oxygen vacancy in the solid solutions were considered as phonon scattering centers. For both solid solutions, the lattice strain effects on the lattice defect thermal resitivities were in preference to the mass effects. In addition, the stoichiometry effects on the additional defect thermal resistivities were about 1.3 times larger than the cation effects.

The thermal conductivities of near-stoichiometric [(U0.8Pu0.2),R]O2 solid solutions containing RO1.5(R = Nd or Eu) up to 10 mol% were determined by the laser flash method in the temperature range 700–1900 K. The thermal conductivities for the solid solutions up to about 1550 K. satisfied the phonon conduction equation K = (A + BT) −1 . The thermal conductivity decreased gradually with the increase of the rare earth content. This decrease was mainly caused by the lattice defect thermal resistance. The measured defect thermal resistivities ( = A ) were in good agreement with the calculated results based on the lattice defect model in which U4+, U5+, Pu4+ and R3+ ions were considered as phonon scattering centers. The lattice strain parameter ϵ = 97 and 103 were obtained for (U, Pu, Nd)O2 and (U, Pu, Eu)O2 solid solutions, respectively. The lattice strain effect on the thermal resistivity was about 15 times larger than the mass difference one.

In-pile release of fission gas from sintered UC pellets in the presence of 8–230 ppm of water vapor in the He sweep gas was measured over the temperature range of 160°–1,000°C. A very complex release behavior was observed and the mechanisms of release were deduced from the manner in which the release depended on the decay constant. It was established that the release of short-lived fission gases during irradiation was controlled mainly by pseudo-recoil, while chemical reaction between UC and water vapor, as well as knock-out, appeared to contribute much more significantly in the case of the longer-lived fission gases. The release of fission gas after reactor shutdown was shown to be governed by the UC-H2O reaction. The ratio of the release due to this reaction in reference to the total release was found to be dependent not only on the concentration of the water vapor but also on the amount present of the accumulated reaction products. Also, a discussion is given on the inordinately high release of 135mXe ...

The thermal conductivities of near-stoichiometric (U, R)O 2 solid solutions (R = Nd, Sm and Eu) containing RO 1.5 up to 15 mol% were determined in the temperature range 700–2000 K by the measurement of thermal diffusivity. The thermal conductivities satisfied the phonon conduction equation K = ( A + BT ) −1 within ± 5%. The constant A corresponding to the lattice defect thermal resistivity increased linearly with the rare earth element content, while the temperature coefficient B was almost independent of it. The change in A with the rare earth element content increased in order of (U, Eu)O 2 , (U, Sm)O 2 and (U, Nd)O 2 solid solutions. The increase of A was explained reasonably by the lattice defect model considering U 4+ , U 5+ and R 3+ ions in the solid solutions as phonon scattering centers, using a common value for the strain parameter ( ϵ = 110). For all solid solutions, the lattice strain effect on the lattice defect thermal resistivities was much larger than the mass effect. In addition, the effect of U 5+ ions on the lattice defect thermal resistivity caused by the lattice strain effect was larger than that of R 3+ ions.

In-pile release of 85mKr, 87Kr, 88Kr, 133Xe, 135Xe and 138Xe from natural graphite was measured over a temperature range of 100 to 950°C and that of 133I and 135I was estimated. The effects of fission rate and fission density on the fission gas release were examined. The β-decay process of iodine within and outside the graphite was observed to control the gross release of xenon. Possible mechanisms of an anomalous sink of 88Kr in the plot of log R B versus log λ were discussed. A new model for knock-out release was proposed on the basis of its temperature dependence found. These results were compared with those of post-irradiation experiments and explained by a mechanism whereby fission gas is trapped in defects created in graphite by fission and β-decay energy, and released through annealing of the defects.

Improvements made on the Fission Gas Release Loop from 1965 to 1970 are reported. These include the development and installation of a cartridge-type specimen chamber, an automatic flow controller and an automatic sampling-measuring equipment, as well as the installation of oxygen and water vapor analyzers. The performance of the safety systems during loop operation are also described.

The kinetics of the oxidations of uranium tetraiodide and of triiodide by oxygen were studied over the temperature ranges from 55.0 to 66.5°C and from 88.5 to 103.5°C respectively. The experiments were carried out with an apparatus in which the column of a gas chromato-graph was replaced by an electric furance and traps of water vapor, oxygen and iodine. Most of the experiments were performed with the same partial pressure of oxygen (64 mmHg) and total flow rate (60 ml. per min.). The activation energies for the reaction of oxygen with uranium tetraiodide and triiodide were, respectively, 21.1 and 22.5 kcal. per mole. The rate of the formation of uranium trioxide was in agreement with the kinetics to be expected for reaction between a gas and a solid at a continuously diminishing spherical interface.

An investigation was undertaken to determine the activation energy of iodine diffusion in graphite and the mechanism of diffusion and formation of the iodide. The heating curve of deposited iodide was found to have a peak at 530 deg C, while a diffusion peak occurred at 750 deg C, with an activation energy of 47 kcal/mole. The results are compared with diffusion of xenon and annealing of graphite irradiated with fast neutrons. (D.L.C.)

Uranium tetraiodide is oxidized to triuranium octoxide through three intermediates, which are deposited on silica powder heated to temperatures of about 400, 300 and 200°C respectively. The intermediates are produced by the action of oxygen contained in small amounts in a helium stream (1/100–1/10000 atm.) on uranium tetraiodide. The intermediates are considered to be compounds consisting of uranium, iodine and oxygen. Chemical analysis, the measurement of the melting point and X-ray analysis show that the compound deposited at the temperature of 400°C is uranium triiodide, while the compounds deposited at 300 and 200°C are both uranium tetraiodides. Three hypotheses have been formulated to interpret the above phenomenon, in which the reaction of uranium iodides with iodine has been studied.

Gas evolution from graphite was studied. The materials used for the study included natural and artificial graphites, two kinds for each, and amorphous carbon, one kind. All the materials were received from the respective makers in the from of powder. Degassing was carried out at 300°, 600° and 1, 000°C in vacuo. The total volume of gases evolved ranged from 410 to 7, 190μl/g of graphite. Oxygen, nitrogen and hydrogenoxide were mainly evolved when degassed at 300°C, and various hydrocarbons up to 50% of the total volume evolved when degassed at 600°C. When degassed at 1, 000°C, carbonmonoxide and hydrogen were predominent among all gases evolved. An analysis of the gases was carried out entirely by gas chromatography. A Molecular Sieve-5A column was used for the inert gas analysis and a two stage column, dioctyl phthalate-A and dimethyl sulforane, for the analysis of hydrocarbons.The total volume and the ratios of the gases evolved from the graphites at experimental temperatures are discussed in detail and the states of existence of the gases in graphite are also discussed.

In-pile release of fission krypton, xenon and iodine from graphite was measured and the release mechanism is discussed in detail compared with the data of post-irradiation experiments. No temperature-independent release of fission gases and iodine was observed in the temperature range from 100 to 640 °C. After reactor shutdown with the specimen temperature constant, only xenon was released from the specimen and not krypton and iodine. To explain this phenomenon, a hypothesis is proposed that fission gas releases are induced by β-decay of precursor halogens trapped in defects in the specimen. By the hypothesis, the release behavior of fission gases and iodine can be explained satisfactorily.