In order to assess the neutron cross section for Mo, the angular neutron spectrum at r=15 cm and μ=0 in a spherical Mo pile of 60 cm I.D. was measured by the linac timeofflight method. The result was compared with the theoretical calculations which were predicted by making use of group constants produced from JENDL-1 (MAT=1420) and ENDF/B-IV (MAT=1287).
The predicted spectrum from the ENDF/B-IV data gives higher flux by about 30% than the measured one in the energy region above 500 keV, while the spectrum calculated from the JENDL-1 data is in better agreement with the measured one.
Substitution of four kinds of partial cross sections from JENDL-1 into ENDF/B-IV was tried and the results were compared with each other. These calculations show that the discrepancy between the measured spectrum and the predicted one from the ENDF/B-IV data is mainly due to the fact that the ENDF/B-IV data has no discrete levels of inelastic scattering of neutrons for Mo.

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Burnout characteristics in a uniformly heated, vertically oriented tube, under flow reduction condition, were experimentally studied. Test pressures ranged 0.53.9 MPa and flow reduction rates 0.635%/s. An analytical method was developed to obtain the local mass velocity during a transient condition. The local mass velocity at the burnout location with an increasing flow reduction rate was slightly different from that measured in steady state tests. The system pressure had a significant effect on the difference. An empirical correlation was presented to give the ratio between the transient and steady state burnout mass velocities at the burnout location as a function of the steam-water density ratio and the flow reduction rate. Experimental results of previous work were compared with this correlation.

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A new, far simpler, method is presented to replace the laborious one-dimensional dif-fusion code calculations currently practised for deriving the axial power distribution through reactor core-an important factor in determining the hot channel factor of PWR cores. With the new method, the axial power distribution is approximated by a combination of simple functions, using as intermediate parameter the "axial offset", representing the extent of imbalance in aggregate power output between the upper and lower core halves.
The new simplified method reduces almost to 1/50 the computer time required for deriving the axial power distribution for any given condition of reactor operation, compared with that demanded by 1-D diffusion code calculations, with deviation of resulting values at peak power position hardly exceeding 2%. The proposed method promises to serve usefully for plant-side calculations and analytical treatment of the axial power distribution.

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Heat transfer by simultaneous conduction and radiation through ceramic fiber insulation is studied experimentally. The temperature profiles and the effective thermal conductivity are determined over a wide range of porosities of the insulation and of temperature pro-vided at the hot boundary.
The resulting data are analyzed in detail to gain an understanding of the mechanism of heat transfer through such medium. A semi-empirical formula is derived for quantita-tively predicting the thermal performance of ceramic fiber insulation. The formula incor-porates as parameters the porosity and temperatures of the hot and cold boundaries.

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Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g•UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum tempera-ture difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g•UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA condi-tions.

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Peak counting rates of a Ge (Li) detector for γ-rays of interest and exposure rates at 1 m above the ground have been calculated using the soil and air concentration profiles of naturally occurring radionuclides at Tsukuba (Yatabe City, Tsukuba District, Ibaraki Prefecture). Decrease in the peak counting rate for 609 keV ²¹⁴Bi γ-ray above the ground, which was due to diffusion loss of ²²²Rn under the ground, was evaluated to be about 44% of the peak counting rate calculated assuming uniform distribution of ²²²Rn under the ground, and decrease in the exposure rate was evaluated to be about 13% of the total exposure rate calculated assuming uniform distribution. Field measurement of i-rays from sources in the soil and air was also made with the Ge (Li) detector at Tsukuba. The calculated peak counting rates where the diffusion loss of ²²²Rn was involved agreed with those observed in situ. During the observation the calculated peak counting rate for the 609 keV γ-ray from airborne ²¹⁴Bi ranged 319% of that from uniformly distributed ²¹⁴Bi source under ground, and the calculated exposure rate from the airborne ²²²Rn daughters ranged 19% of the total exposure rate.

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In order to relieve the difficulties of repair and maintenance and to make the reactor size compact, a concept of tokamak reactor which is installed in a water pool has been proposed. Preliminary design study of the concept was carried out. As the result of this study the following advantages over conventional tokamak reactors are shown ; The size of TF coil can be considerably reduced while retaining sufficient space for repair and maintenance because a solid shield is eliminated. Disassembling and reassembling of vacuum vessel seems to be done with realistic remote handling technique. The problem caused by radiation streaming can be considerably eased. Radioactive waste disposal is reduced con-siderably because a solid shield is eliminated.

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