Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident.

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A new homogenization technique for neutron cross section in the heterogeneous region has been developed for shielding transport calculations. The proposed homogenization technique consists of two steps. The first step is the flux weighting method, in which averaged cross sections are produced so as to conserve the reaction rates using the neutron flux obtained from the heterogeneous cell calculation. In the second step, the averaged cross section values obtained from the first step are adjusted in order that the gradient of the neutron flux in the homogeneous calculation using the adjusted cross section will agree with that in the heterogeneous cell calculation.
Calculated results of three homogenization techniques have been compared with that of heterogeneous calculation in the case of an LMFBR B₄C axial shield. The volume weighting method, which is usually used in shielding design calculations, underestimates the total neutron flux by a factor of 4.0.
The flux weighting method, although providing a better approximation than the volume weighting method, still underestimates by a factor of 1.7. On the other hand, the proposed homogenization method shows good agreement with the heterogeneous calculation within an error of few percent. The proposed homogenization technique is shown to be effective in improving the accuracy of practical shielding transport calculations.

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A reactivity effect due to a spatial variation of nuclear fuel concentration is an important problem for nuclear criticality safety in a reprocessing plant. As a theory estimating this reactivity effect, the Goertzel and fuel importance theories are well known. It has been shown that the Goertzel's theory is valid in the range of our experiments based on measurements of reactivity effect and thermal neutron flux in non-uniformly distributed fuel systems. On the other hand, there have been no reports concerning systematic experimental studies on the flatness of fuel importance which is a more general index than the Goertzel's theory.
It is derived from the perturbation theory that the fuel importance is proportional to the reactivity change resulting from a change of small amount of fuel mass. Using a uniform and three kinds of nonuniform fuel systems consisting of 93.2% enriched uranium plates and polyethylene plates, the fuel importance distributions were measured. As a result, it was found experimentally that the fuel importance distribution became flat, as its reactivity effect became large. Therefore it was concluded that the flatness of fuel importance distribution is the useful index for estimating reactivity effect of non-uniformly distributed fuel system.

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In a charge-division type position-sensitive proportional counter (PSPC) with an anode wire of small resistance, a reflected component from an opposite end and thermal noise involved in signals deteriorate the position resolution of the PSPC. A digital waveform processing method was applied to the reduction of these undesirable effects by skillfully utilizing their signal characteristics that can be observed as inversely correlative signals between two-output signals from both sides of the PSPC. The digital waveform processing could improve the position resolution compared to a conventional pulse height processing method with analog filters. When the digital waveform processing was applied to signals of an equivalent circuit simulating the PSPC, the position resolutions defined by the full width at half maximum were improved to about 30% of those of conventional analog pulse processing. In the case of an actual PSPC, the position resolutions by the digital waveform processing were improved by 4-10% as compared with those of conventional pulse height processing.

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An evaluation method for droplet deposition coefficients, corresponding to spacer shapes, was developed in order to predict nuclear fuel bundle critical power. The method is based on a three-dimensional numerical analysis of dispersed flow around the spacer. The analysis showed that coolant droplet deposition characteristics onto fuel rods differed between grid type and ferrule type spacers, and also between fuel rods due to flow tabs on the upper end of the spacer band. The droplet deposition coefficient was modeled from the analysis results and adopted into a critical power prediction code based on subchannel analysis. Comparisons between critical powers calculated by the subchannel analysis code and those measured with full-scale test assemblies showed good agreement and improved prediction accuracy of the code.

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A behavior of waterside corroded PWR fuels under transient conditions was studied. In order to simulate the corroded condition, unpressurized test fuel rods having cladding oxidation up to 80μm and hydrogen contents up to 344ppm in maximum were fabricated and pulse irradiated in the Nuclear Safety Research Reactor (NSRR) belonged to the Japan Atomic Energy Research Institute.
Major results obtained are:
(1) The waterside corroded fuel rods having oxides to the magnitude of 50, 60, 70 and 80μm did not fail in the energy deposition range from 256 to 279 cal/g•fuel. The value was exceeded the failure threshold of 260cal/g•fuel for NSRR standard fuel. While, the reference (no oxide) fuel rods in the same energy deposition range failed by melt/brittle of the cladding followed by wrinkle deformation.
(2) From posttransient irradiation examination it was found that wrinkle deformation in the waterside corroded fuels occurred very little due to rather relatively low cladding surface temperature and small diametral residual strain than those in the reference fuels.

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Oxidation tests of irradiated and un-irradiated rod with an artificial defect hole were performed in air and air-argon mixed gas between 473K and 513K to study the oxidation behavior of damaged fuel rod during dry storage. Rods oxidized in air showed diameter increase due to volume expansion caused by the generation of U₃O₈ and the cladding of irradiated rod was broken after the diameter increase had reached about 2%. X-ray diffraction and metallography showed that the oxidation of irradiated UO₂ pellet occurred preferentially at the grain boundary and that the crystal structure corresponding to U₄O₉ remained stable even after about 11, 200h oxidation in air at 473K. It was suggested that the difference of the oxidation between irradiated and un-irradiated rods was due to FP gas bubble accumulation at grain boundary and FP accumulation in UO₂ matrix.
The oxidation rate of fuel rod in air-argon mixed gas was lower than that in air irrespective of irradiation. The irradiated rod in low oxygen partial pressure showed very little deformation.

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A comparative study of the shielded area space required for storing fissile solution by the conventional annular tank and by poison tube tank is made. Poison tube tank is similar to commercial heat exchanger. The neutron poisons studied are gadolinium oxide and borax. Variation of multiplication factor for an array of annular tanks containing uranium nitrate or plutonium nitrate solutions are presented for annular widths of 10, 7.5 and 5cm. It is concluded that for the given concentration, 5cm annular width tanks are safe at a pitch distance of 120 and 90cm for uranium and plutonium solutions respectively. Using these, as reference values, it is found that the shielded area saving for the poison tube tank is a factor of 12 and 8 for the given concentration of uranium and plutonium solutions respectively.

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If the core power inherently follows change in heat removal rate from the primary coolant system within small thermal expansion of the coolant which can be absorbed in a practical size of pressurizer, reactor systems may have more safety and load following capability. In order to know possibility and necessary conditions of a concept on reactor core and primary coolant system of a pressurized water reactor (PWR) with such"highly inherent heat removal following capability", transient analyses on an ordinary two-loop PWR have been performed for a transient due to 50% change in heat removal with the RETRAN-02 code.
The posibility of a PWR concept with the highly inherent heat removal following capability has been demonstrated under the conditions of the absolute value of ratio of the coolant density reactivity coefficient to the Doppler reactivity coefficient more than 10×10³kg•cm³ which is two to three times larger than that at beginning of cycle (BOC) in an ordinary PWR and realized by elimination of the chemical shim, the 12% lower average linear heat generation rate of 17.9 kW/m, and the 1.5 times larger pressurizer volume than those of the ordinary PWR.

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The oxidation of Zircaloy-4 in steam at 1, 173 to 1, 773K has been modeled by solving the oxygen diffusion equation numerically to predict the oxidation rate of the Zircaloy-4 at high temperature. A fully implicit finite difference method was employed to solve the diffusion equation for a one dimensional cylindrical geometry. The influence of the heat flux on the isothermal oxidation rate has been considered to show its effect on the oxide-metal interface temperature and the acceleration of the oxidation rate. The results obtained were found to be in good agreement with experimental data reported in the literature.

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A computer aided process flowsheet design and analysis system, COMPAS has been developed in order to carry out the flowsheet calculation on the process flow diagram of nuclear fuel reprocessing. All of equipments, such as dissolver, mixer-settler, and so on, in the process flowsheet diagram are graphically visualized as icon on a bitmap display of UNIX workstation. Drawing of a flowsheet can be carried out easily by the mouse operation. Not only a published numerical simulation code but also a user's original one can be used on the COMPAS. Specifications of the equipment and the concentration of components in the stream displayed as tables can be edited by a computer user. Results of calculation can be also displayed graphically. Two examples show that the COMPAS is applicable to decide operating conditions of Purex process and to analyze extraction behavior in a mixer-settler extractor.

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