To provide accurate double differential neutron emission cross sections of fusion reactor candidate elements at En = 14 MeV, a fixed 8.3 m TOF facility with an NE213 detector of 25.4 cm diameter was installed in the 85° direction of a pulsed D-T source. Detail of scattering experiment using this facility is described. Energy resolution is ± 0.21 MeV at 14 MeV and good counting statistics is attained by about 1/h run using a cylindrical sample of 3 cm in diameter and 7 cm long.
Measurements of double differential neutron emission cross sections are carried out for C, F, Mg, Al, Si, V, Cu and Pb. Results are compared with evaluated neutron emission data. Energy resolution is sufficient enough to deduce partial angular differential cross sections for elastic and discrete inelastic scatterings.

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A numerical technique for solving liquid metal MHD problems has been developed as a design tool of fusion reactor blankets. Blanket cooling ducts have spatial changes in boundary conditions of flow. The changes generate recirculating currents in the fluid which affect the velocity profile. The present code can analyze the MHD problems in the ducts which have planes of symmetry perpendicular to the uniform magnetic fields. One example is a straight circular pipe with axially varying wall conductance. Under the very strong uniform field, the current etc. in the fluid are derived to be uniform in the field direction. Therefore, these values in the diametric plane perpendicular to the field well represent the fluid properties in the whole duct. Thus we can replace the three-dimensional problem with the two-dimensional one. The following calculational procedure is used in the code. With an assumed velocity distribution, the electric potential and current distributions are calculated at first using a finite difference method. The result usually does not satisfy all the governing equations. The velocity is redistributed iteratively until the solution which satisfies all the governing equations is obtained.

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The onset conditions of vortex-induced air entrainment were investigated for the suction flow into a vertical drawing pipe in a cylindrical test vessel. It was found that the onset conditions can be represented by the following two types of empirical criteria :
where D is the inner diameter of the suction pipe in the unit of meter, H the submergence, Fr the Froude number (= V12/gH), and Ro the rotational Froude number (= C²_∞/gH³). The subscript n on Ro and Fr denotes the onset correlations derived independently from the relation between Ro and H/D and the relation between Fr and H/D. From visual observation and theoretical consideration, it was found that the above two criteria come from the fact that the onset conditions are dominated by the following two factors : whether the air core formed by vortex reaches the pipe, and whether the water flow in the suction pipe is fast enough for the flow to draw bubbles downward. By combining the two onset criteria into single correlation and comparing it with experimental data, the following correlation was derived,
It was confirmed that this correlation represents all the experimental data well.

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Natural convection in a hemisphere was studied with bottom surface heated and hemispherical surface cooled. Heat transfer and flow visualization experiments were performed. The range of Rayleigh number and Prandtl number were 9×10⁶ <Ra <7×10⁹ and 6 <Pr <860 for heat transfer experiment and 2×10⁵ <Ra×10⁷ and 6 <Pr< 860 for flow visualization experiment. Heat transfer characteristic was empirically correlated by the following relation as a function of Rayleigh number : Nu=0.2352 Ra0.242. The value of the exponent 0.242 is close to 0.25 which corresponds to the case of the laminar natural convection in vertical parallel plates. The measurement of the fluid temperature in vertical direction showed that hot thermal plumes reached near the top of the hemisphere for low Rayleigh number (Ra <1×10⁹) and the hot thermal plumes played an important role on the liquid temperature reversal observed for low Rayleigh number. For high Rayleigh number (Ra >1x 10⁹), the effect of the hot thermal plumes was confined in the vicinity of the hot surface and the temperature reversal disappeared. Flow visualization experiment revealed the existence of downward flow along the inner hemispherical surface. The existence of the downward flow suppresses the generation of the cold thermal plumes.

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There has been increasing necessity for load following and/or AFC (Automatic Frequency Control) operation along with the growth in the share of nuclear power generation in the electric power network. Fuzzy logic control was investigated for application to a BWR recirculation flow control system, in order to obtain a rapid generator power response within an allowable neutron flux overshoot. The proposed controller has two control loops, generator power and neutron flux loop. The fuzzy logic is utilized for weighing these control loops and for controlling the neutron flux. By evaluating the controller performance by numerical simulations on the step response for generator power demand with the model BWR recirculation flow system, more rapid response was obtained than that for conventional proportional plus integral controllers with no neutron flux overshoot beyond alarm activation level.

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In the previous study, it is reported that the core collapsed liquid level was depressed nearly to the core bottom and the dryout of the core was observed in the early stage of the PWR cold leg small break loss-of-coolant accident (LOCA) experiment, The manometric effect due to the liquid seal formation in the loop seal and the difference of the liquid holdup between the steam generator (SG) upflow-side and downflow-side caused a depression of the core collapsed liquid level. The core liquid level was recovered just after the loop seal was cleared.
The bypass between the core side and the downcomer side affects the core liquid depression. Four 5% cold leg break experiments with the different core bypass location, configuration and size were conducted to clarify the bypass effect. When the bypass was relatively small (less than 3% bypass of the initial core flow before the break), the timing of the loop seal clearing delayed with the bypass. When the bypass was relatively large (9, 2% of the core flow), the loop seal clearing took place after the break uncovery and the timing was significantly delayed, In general, the smaller minimum core collapsed liquid level was obtained at the earlier timing of loop seal clearing due to the smaller bypass.

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The experiments on the migration of ¹³⁷Cs adsorbed on fine soil particles in unsaturated sandy soil layers were carried out. The soil layer reduced the concentration of ¹³⁷Cs in the effluent, but did not retard its migration. The concentration of ¹³⁷Cs remained in the soil layer increased in proportion to the amount of ¹³⁷CS introduced.
By the measurement of size distribution of fine soil particles, it was revealed that the concentration of ¹³⁷Cs in the effluent decreased exponentially with the length of layer in the case of monodispersed fine soil particles.
From these results, the first-order irreversible reaction formula was obtained in order to formulate the reaction of ¹³⁷Cs adsorbed on fine soil particles with soil layer. A good agreement was obtained between the predicted result by polydispersed irreversible model and experimental one.

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The effect of background counts induced by a-particles mainly from Rn in air was removed by the attachment of a water-permselective membrane separator to a tritium stack monitor equipped previously. Water vapor, after permeating the membrane selectively, was carried by N<TUB>2</TUB> gas into an ionization chamber where the activity of tritium was measured. The consumption of N<TUB>2</TUB> gas for carrier was reduced by recycling the gas through dehumidification processes ; (1) condensation by pressure, (2) condensation by refrigeration and (3) adsorption (by adsorbents), so that there is no added daily work arising from the attachment for maintenance of the tritium monitor.

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The safety research for BWRs has been positively done by the JAERI, Japanese BWR utilities and BWR vendors in this decade and has shown the important phenomena under BWR LOCA conditions. Based on these significant results, the SAFER03 computer code was jointly developed by Toshiba, Hitachi and General Electric. SAFER03 has been qualified against the BWR simulation test data obtained from TBL, ROSA-III and FIST-ABWR test facilities. The objectives of this study are to assess the predictive capability of SAFER03 code to simulate the significant LOCA phenomena and to catch key parameters during BWR LOCA. This paper summarized the results of these SAFER03 assessments and showed that SAFER03 could predict the realistic behavior of BWR LOCA with slight conservative peak cladding temperatures.

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