The release rate of fission-gas from UO₂ was continuously measured during irradiation in the Hitachi Training Reactor. The UO₂ specimen was heated electrically in in-core assemblies by tungsten heaters, either arranged axially transversing the specimen (producing radial temperature gradient) or cylindrically outside the specimen (uniform heating). In a case of the axially heated annular UO₂ pellets with radial temperature gradient, the rate of fission-gas release increased with time under constant temperature and flux. This experimental result was explained by considering as fission gas release mechanism a combination of diffusion and pore migration processes. On the other hand, when UO₂ pellets were heated isothermally up to 2, 000°C, the fission-gas release rate could be explained in terms of diffusion process alone. No effect of equiaxial grain growth was observed on the rate of fission-gas release.

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A method of modal expansion approximation is applied to study a burnup optimization problem. The spatial distribution of the neutron flux is approximated by a linear combination of certain predetermined spatial modes, and one of these modes is regarded as the control mode. A computational procedure that allows fast and sufficiently accurate estimation of the effect of flux shaping on the attainable burnup is described. As numerical example the optimal policy for flux shaping for a one-dimensional slab reactor model with nonlinear feedback effects is sought by this method. By manipulating the flux shape according to the optimal policy, the attainable burnup is increased appreciably over that obtained by the conventional method based on constant flux distribution, when the maximum allowable power peaking factor is large. The optimal policies are determined uniquely in the cases of highly nonuniform fuel loading and smaller values of the maximum allowable power peaking factor, while they become non-unique in the contrary case. The present method is applicable to more general problems such as the optimization of flux shaping of a reactor with multi-zone refueling scheme without much increase in computing effort.

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Uranium nitride and uranium carbonitride fuel pellets were prepared for irradiation in the Japan Material Testing Reactor. The pellets are 6.9mm in diameter and 7mm long, and are of natural and 5% enriched uranium. Uranium nitride powder was prepared from uranium metal via hydride and higher nitride. Uranium carbide powder was prepared from uranium metal by hydriding and then reacting with propane. The lowest possible reaction temperatures were selected to obtain fine and reactive powders. Uranium nitride and mixed powders of different ratios (UC:UN=1:3, 1:1 and 3:1) were cold pressed without binder. Sintering was carried out in a tungsten crucible in vacuum (10^-4mmHg) for 2hr at 1, 900°∼2, 000°C. The density of the pellets obtained was in the range of 90∼95% of the theoretical value with an oxygen conten of 1, 300∼2, 100ppm. No second phase, such as metallic uranium, were observed in the specimens, either by metallography or X-ray diffraction. These pellets of unexpectedly high density without second phase must have been obtained thanks to the good powder characteristics combined with proper sintering conditions. The compositions of uranium carbonitride pellets were found to be slightly nitrogen deficient, compared with the reactants.

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The modified damage constant. K_L=(1/L_i²-1/L₀²)•Φ^-1, is obtained from the drop rate of the photocurrent at 1.0μm monochromatic light irradiation, where the minority carrier diffusion length in the bulk region before γ-irradiation, L₀, is estimated by fitting the observed spectral response carve to those calculated with the diffusion length as parameter. The values L_i and Φ represent the minority carrier diffusion length after γ-irradiation, and the total flux of incident γ-ray photons, respectively. When N/P-type cells are contaminated with Cu, the values of K_L by about one order of magnitude compared with non-doped cells, while those of the P/N-type cell contaminated with either Cu or Ni ane only slightly smaller than when not doped. The K_L values of the pulled (C•Z) bulk P/N-type cell are about 1/2 those of the floating zone (F•Z) bulk cell. The curves of K_L of the non-doped F•Z and C•Z bulk P/N-type cell vs. total dosage begin to decrease from a point below 10¹⁶ photons/cm² total dose. Also, the damage constant, K=(τ_i^-1-τ₀^-1)•Φ^-1, of n-type floating zone (F•Z) and pulled (C•Z) bulk P/N-type cells and of p-type C•Z bulk N/P-type cells, both γ-irradiated, is seen to increase with dopant concentration in the bulk region.

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The thermal decomposition of PuF₆ in the temperature range from 100° to 160°C was studied. At 100° and 120°C, no decomposition was detected. At 140° and 160°C, the rate of decomposition during reaction lasting up to 20hr, and the dependence of the decomposition on the initial PuF₆ pressure were determined. The results are explained by assuming that the rate of decomposition is proportional to the 0.4-th power of PuF₆ pressure. The activation energy of the reaction was found from calculation to be 14.8kcal/mole.

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Thermo- and photoluminescence were studied on neat 2-methyltetrahydrofuran glass after γ-irradiation at 77°K in dark. When the glass was warmed, thermoluminescence was observed at about 97°K. Photoluminescence was also induced by light of λ>600nm. The luminescence spectrum was structureless with a maximum at 460nm, which coincides with the luminescence induced by stimulating free radicals with ultraviolet light (270∼350nm). Both the thermo- and photoluminescence are interpreted as due to the formation of an excited free radical by the charge recombination reaction, Lifetime measurements indicate that the free radical is probably excited to its fluorescent state.

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The polarity correlation method has been applied to an experiment for measuring the prompt mode neutron decay constant of a reactor with a long neutron lifetime. Measurements have been performed in the SHE VIII-1 at critical and several subcritical steady states (0∼-1$).
The analog signal of detector output is converted to a polarity signal having only one-bit of information on the amplitude of the detected signal. The polarity correlation function is obtained as the correlation function between two polarity signals. Attention has been paid in the design of the correlator to obtain ample stability in operation. Using a digital computer, the prompt mode neutron decay constant is determined with 2% experimental accuracy from the decay curve of the polarity correlation function. A series of the prompt mode neutron decay constants is fitted to an approximated expression of inhour equation to obtain β/l. The value of each prompt mode neutron decay constant was found to be in good agreement with that measured by the pulsed neutron technique. The polarity correlation method is particularly applicable to critical or near critical state where a pulsed neutron source cannot be effectively introduced.
In Appendix, an estimation is made for the error in the measured polarity correlation function due to unbalanced mean setting level in the experiment.

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A probabilistic model for radiation transport calculation is presented based on the assumption that the motion of particles can be sampled at predetermined points. Due to the fixed sampling points, the directions of motion of particles are discrete however the energy of particles can be followed continuously. More specifically the transport medium is considered to be filled with cubic lattice and sampling is made at the corner of the cubes.
A brief description of Boltzmann transport equation and its connection to the mathematical development of this model is clearly shown. The model and its development is general and has application in many fields associated with radiation transport calculation.

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