The total cross section of beryllium in the 0.002-0.3 eV neutron energy range was measured by transmission technique at four different temperatures of the sample-300°, 573°, 773° and 973°K.
The scattering kernel was directly calculated from the frequency distribution function, instead of using S(α, β) as practiced in the GASKET-FLANGE method. Integration of the scattering kernel based on the Sinclair model, the Young-Koppel model and the Raubenheimer-Gilat model resulted in little difference between the calculated total cross sections obtained with the three models.
The experimental values are in good agreement with calculations for all measured temperatures of the sample.

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The velocity of a single bubble in quiescent and flowing liquid was studied to gain information on the structure of bubble- and slug-flows of gas-liquid two-phase flow.
In the experiments, tap water was used at room temperature and the bubbles were generated by injecting air with a syringe. The bubble velocities were evaluated from photographs taken by multiflash exposure. The tube diameters adopted were 5, 1 and 0.5 cm for the flowing liquid experiments and 1, 0.5 and 0.2 cm for the quiescent liquid experiments. The average liquid velocities varied from about 0 to 2 m/sec.
The results indicated that the velocity of a single bubble in flowing liquid was the sum of the local liquid velocity in the vicinity of the bubble and of the velocity of rise of bubble in quiescent liquid. For Taylor bubbles, the local liquid velocity is given by the maximum liquid velocity near the center of the channel.
The velocities of spherical and ellipsoidal bubbles are determined by the balance of forces acting on them, but that of the Taylor bubble appears to be determined by a different mechanism. A good explanation of such a mechanism is Taylor instability at the gas-liquid interface well removed from the channel wall.

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An experimental study was performed on the heat transfer to a turbulent water film which is falling down the outside surface of a heated vertical tube. The test section was made of stainless steel tube 13 mm in outside diameter and 1, 500 mm long. It was found that a nearly constant heat transfer coefficient was obtained for the lower half of the test section covering a length of about 1 m, which was direct-heated by electric current. The resulting values of the Nusselt number in this region were correlated to the Reynolds number, and the plots fell in a region intermediate between those given by Wilke and by McAdams. An analytical model taking account of the undulation on the film surface is proposed. The predictions from the model give fairly good agreement with the experimental data.

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Thermal diffusion constants were determined over a temperature range of 400°700°C in liquid sodium alloyed with trace amounts of zinc, antimony or cesium.
Purified sodium containing a radioisotope of one of the three solute metals was filled by suction in a capillary made of stainless steel. The capillary was then carefully seal-welded to prevent the sodium from oxidation; the sealed capillary was placed along the core of a heat conductor rod; a steady linear temperature gradient was produced along its length by heating its head and cooling its foot. After a predetermined period of this thermal diffusion run, the capillary was quickly chilled to solidify the sodium and sectioned into pieces.
The specific radioactivity observed in the sodium in the sectioned piece were plotted in log-log scale against the absolute temperature prevailing in the piece during the thermal diffusion run, and the thermal diffusion constant was obtained from the inclination of the straight lines drawn through the plots. All three solute metals were found to have become enriched in the lower temperature region. The measured thermal diffusion constants were respectively 2.1 for zinc, 1.7 for antimony and 1.6 for cesium.

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Internal friction measurements were performed on neutron-irradiated and annealed Fe-0.2w/o Cu-0.0066 w/oC alloy, in an attempt to gain information on the role of interstitial C atoms in radiation effects. It was revealed that C atoms are trapped by defects produced by neutron-irradiation, and released upon annealing at about 400°C. The radiation hardening induced in the neutron-irradiated Fe-Cu-C alloy recovers by annealing in at least three steps of 150°250°C, 350°450°C and above 550°C. The second step corresponds to the release of C atoms from traps.

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The thermal diffusion of hydrogen is one of causes of uneven hydride precipitation in zircaloy fuel cladding tubes that are used in water reactors. In the diffusion model of hydrogen in zircaloy, the effects of the hydride on the diffusibility of hydrogen has been regarded as negligibly small in comparison with that of hydrogen dissolved in the matrix. Contrary to the indications given by this model, phenomena are often encountered that cannot be explained unless hydride platelets have considerable ostensible diffusibility in zircaloy.
In order to determine quantitatively the diffusion characteristics of hydrogen in zircaloy, a thermal diffusion experiment was performed with zircaloy-2 fuel cladding tubes containing hydrogen beyond the terminal solid solubility. In this experiment, a temperature difference of 20°30°C was applied between the inside and outside surfaces of the specimen in a thermal simulator.
To explain the experimental results, a modified diffusion model is presented, in which the effects of stress are introduced into Markowitz's model with the diffusion of hydrogen in the hydride taken into account. The diffusion equation derived from this model can be written in a form that ostensibly represents direct diffusion of hydride in zircaloy. The apparent diffusion characteristics of the hydride at around 300°C are D_p=2.3 x 10⁵exp(-32, 000/RT), (where R: gas constant, T: temperature) and the apparent heat of transport Q_p=-60, 000 cal/mol. The modified diffusion model well explains the experimental results in such respects as reaches a steady state after several hours.

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Local power peaking was measured by γ-scanning in BWR type fuel assemblies under cold clean conditions using the zero-power critical facility TCA. The horizontal power distribution in an assembly was examined on several loading patterns made up of 1.9, 2.1, 2.3 and 2.6% ²³⁵U enriched UO₂ fuel rods. The power depression was correlated with the lowered value of ²³⁵U enrichment in a fuel rod loaded in the assembly to replace one of normal enrichment. The results show that the local power peaking factor can easily be depressed down to 1.3 by suitable arrangement of depleted fuel rods in the assembly, while more elaborate manipulation is required to reduce it below 1.2.

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