A new concept is presented for direct enrichment of fissile fuel in the blanket of a fusion-fission hybrid reactor. The enriched fuel produced by this means can be used in fission reactors without reprocessing. The outstanding feature of the concept is the powdered form in which UO₂ fuel is placed in the reactor blanket, where it is irradiated to the requisite enrichment for use as fuel in burner reactor, e.g. 3%. After removal from blanket, the powder is mixed to homogenize the enrichment. Fuel pellets and assemblies are then fabri-cated from the powder without reprocessing. The concept of irradiating UO₂ in powder eliminates the problems of spatial nonuniformity in fissile enrichment, and of radiation damage to fuel clad, encountered in attempting to enrich prefabricated fuel. Powder mixing for homogenization brings the additional benefit of removing volatile fission products. Also burnable poison can be added, as necessary, after irradiation.
An extensive neutronic parameter survey showed that the optimum blanket arrangement for this enrichment concept is one presenting a fission suppressing configuration and with beryllium adopted as moderator. By this arrangement, the average ²³⁹Pu enrichment obtained on the natural UO₂ fuel in the blanket reaches 3% after only 0.56 MW•yr/m² exposure. A conceptual design is presented of the blanket, together with associated fusion breeder, from which, practical application of the concept is shown to be promising.

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The scalar neutron spectrum normalized in reference to unit source neutron was meas-ured with a miniature NE213 spectrometer at several positions along the centerline in a graphite pile irradiated with D-T neutrons. The measured spectra were compared with the results of calculation using the MORSE-GG Monte Carlo code with the modified point-detecter estimator of Carter & Cashwell and the GICXFNS group cross section set derived from the ENDF/B-V library. The measured and calculated spectra agreed within the errors estimated from the observed ranges of statistical error and of the folds on spectra due to response function error, except near the resonance peaks of the carbon total cross section at the deeper positions inside pile.

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The CACE (Compressible Advanced Continuous-Fluid Eulerian) method, a numerical method for analyses of RIA (Reactivity Initiated Accident) induced coolant dynamics is ex-amined. It is a semi-implicit method whereby an energy equation plays an important role in convergence of iterations and in calculating the pressure transient. A hypothetical BWR rod drop accident, at a hot stand-by condition, is calculated to evaluate the method applica-bility to RIA induced coolant dynamic analyses using a code composed of one-point kinetic and one-dimensional multi-channel models. The calculated results demonstrate successful application of the CACE method. In addition, comparison with results by RELAP4/MOD6 confirms that the CACE method is superior in terms of stability. It can calculate the pres-sure change caused by rapid heating, a problem which cannot be calculated for the same time step width by RELAP4/MOD6.

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The differences in the single-phase forced-convection heat transfer characteristics between upflow and downflow were investigated experimentally with a narrow vertical rectangular channel. The objectives of the experiment were to investigate in both laminar and turbulent flow regions the applicability of existing correlations to and the effects of buoyant force on the heat transfer characteristics in the narrow vertical rectangular channel, which is simulat-ing a subchannel of 2.25 mm in gap and 750 mm in length in the fuel element of the research reactor, JRR-3 to be upgraded at 20 MWt. As the results, it was revealed that (1) by use of equivalent hydraulic diameter, existing correlations are applicable to a channel as narrow as 2.25 mm in gap for turbulent flow though the precision and critical Reynolds number are different among the correlations, and (2) in the laminar flow, the difference in heat transfer characteristics arises between upflow and downflow with Reynolds number less than about 700 and Grashof number larger than about 1, 000, giving smaller Nusselt number for down-flow than for upflow as the effect of buoyant force. New heat transfer correlations for channel heated from both sides are proposed as lower limits for upflow and downflow, respec-tively, in the laminar flow.

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Two simulation tests of a BWR loss-of-coolant accident (LOCA) by a postulated guillotine rupture of a recirculation suction line were conducted using the Two Bundle Loop (TBL), which was volumetrically scaled to a BWR/5 plant with 764 fuel bundles. The major objective of the tests was to clarify thermal-hydraulic difference in parallel bundles. In the tests, the failure of a diesel generator for two low pressure coolant injection (LPCI) pumps was as-sumed, and the initial bundle power combinations were 4.0 and 5.9 MW in the first test, and 5.0 and 4.9 MW in the second.
In one of the two bundles, the rods heated up locally in the radial direction. In the other, the rods heated up rather uniformly and later than in the former bundle. Much water fell locally into the former bundle, while ascending steam flow from the lower plenum was larger in the latter. A difference in thermal-hydraulic responses was observed even in the case of nearly identical bundle powers, but the difference was less than in the case of dif-ferent bundle powers. The peak cladding temperatures in the tests were lower L han 704 K.

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Thermal conductivity, electrical resistivity and stored energy were measured for Pechiney nuclear grade graphite irradiated in the temperature range 220400°C up to the maximum neutron fluence 2.2×10²⁰ n/cm² (5>0.85 MeV) in the environment of a carbon dioxide in a commercial reactor. Thermal conductivity decreased, electrical resistivity and stored energy increased owing to neutron irradiation and their changes were larger for the samples irradi-ated at lower temperatures.
A linear relation between stored energy and fractional change in thermal resistivity was obtained for the irradiated samples and it was found that its proportional constant is about two times of that reported previously. The relation between thermal conductivity and elec-trical resistivity is discussed for irradiated samples as well.

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Boron-10 lined proportional counters for high temperature use were fabricated and tested. A mixture of Ar and CO₂ was found to be a better ionization gas than that of Ar and N₂ in terms of stability at high temperature. Counters filled with an Ar+CO₂ mixture operated satisfactorily up to 200°C and remained stable over 2 months of operation at 200°C. The change of counter characteristics due to temperature variations and a continuous operation at 200°C were small ; it was ascertained that the counters can be operated in such a way that the output count rate exhibits no significant change for a given neutron flux under exposure to temperature changes or a continuous operation at 200°C.
The counters are intended to be used for the fuel loading channel of the Nuclear Instru-mentation System of "Monju" (Japanese Prototype Fast Breeder Reactor).

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