The Reich-Moore approximation of the R-matrix theory was applied to the analysis of selected measurements of neutron effective total cross sections, fission cross sections and capture cross sections of 2331J in the energy range from thermal to 150 eV. The resonance parameters were obtained by fitting the experimental data by the Bayesian code SAMMY. Results of the calculated cross sections are compared with the corresponding experimental data in graphs and tables. The statistical properties of the resonance parameters were examined and the average parameters were obtained. The resonance parameters are given in an ENDF-6 format file available from the JAERI Nuclear Data Center and from the NEA Data Bank (OECD, Paris).

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Identification of radionuclides and evaluation of dose rate level around a vacuum vessel of
JT-60U tokamak device have been made. A one-dimensional calculation code was employed in this work. Radionuclides of ⁵⁶Mn (High-Mn steel of toroidal field coil case), ⁵⁸Co (Inconel-625
vessel) and ⁶⁰Co (SS-316 material of first wall) appeared after shutdown of operation with deuterium gases. Cobalt-58 and ⁶⁰Co with long half-life time contribute to the dose rate on the vessel by 7 to 3 proportion. The one-dimensional calculation provided a sufficient prediction for the dose rate on Inconel vessel of JT-60U, in which the activation property is mainly determined by ⁵⁸Co generated via threshold neutron reactions. The results calculated for the vessel dose rate showed a good agreement with data measured for the first 80 D-D operation weeks.

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The fuzzy logic controller was developed to control load-follow operations in pressurized water reactors. The reactor core characteristics change according to different fuel cycles or core exposures, thus making a nonlinear time-varying control problem. This proposed method, however, does not require a mathematical model to design the controller, and so avoids redesigning or tuning controller gain for various cores. Clearly, this method is very suitable for reactor load-following operation control. The control system has two subsystems: one is to track the desired power, and the other is to keep axial offset close to the target value. Both controllers use fuzzy logic: one is the conventional type, and the other uses fuzzy logic to tune the parameters of the controller so the controller can correspond to various core characteristics. Simulation results show that the control system performs well for different cores, and so this system is useful for load-follow operation.

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A nonlinear reactor dynamics model of reduced order is derived and an analytical study on
BWR power oscillation is made using this model. It provides some essential features which are not given by numerical studies, such as the explicit expressions of the linear stability condition and the weak stability condition which is related to the periodic motion. In addition, the relation between the reactivity feedback and these conditions is obtained. The application of the analytical results to the qualitative analysis of BWR dynamics is easy and quick in comparison with numerical approaches.

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Dissolution of UO₂ crucibles by molten Zircaloy-4 (Zry) was investigated in the temperature range of 2, 223-2, 373 K and for specimens having UO₂/Zry mole ratios between 7 and 18.2.
The uranium concentration in the Zry melt rapidly increased during initial reaction time and approached saturated values, depending on reaction temperature and UO₂/Zry mole ratio.
Kinetics of uranium concentration increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The saturated uranium concentration in the Zry melt was inversely proportional to the UO₂/Zry mole ratio. An empirical correlation of saturated uranium concentration in the Zry melt was obtained as a function of UO₂/Zry mole ratios and reaction temperature. This study of the empirical correlation was intended to estimate maximum UO₂ fuel dissolution by molten Zry cladding during severe fuel damage accidents for three different reactor type fuels.

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General corrosion tests of Type 304SS, Alloy X750 and Stellite #6 used for BWR primary component materials were performed under various conditions of simulated BWR primary water.
The authors' interest was focused on the corrosion behavior, such as metal release and the buildup of corrosion oxide film. It was possible to recognize these corrosion behaviors from the nature of the oxide film of which morphology and property depended on the tested corrosion environment;
(1) The metal release of chromium and iron, the elemental composition of the corrosion oxide films, and electrochemical potentials (ECPs) depended on the oxygen concentration in the test environment.
(2) The corrosion rate of Type 304SS showed its minimum value in a passive region around
0 mV vs standard hydrogen electrode (SHE). It increased slightly by γ-ray irradiation for normal water chemistry (NWC) and it became 2 times for hydrogen water chemistry
(HWC).
(3) The corrosion rates of Alloy X750 and Stellite #6 were over 4 times that of Type 304SS
in oxygenated environments such as in-core water containing H₂O₂.
(4) The differences in the corrosion rates among these alloys were due to the nature of the oxides determined from the composition of the base alloys and the tested corrosion environments.

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Behavior of Tc(VII) in the codecontamination process of the Purex was analyzed using various flowsheet conditions with the EXTRA-M code, which includes a numerical model of distsibution ratio of Tc(VII). The enhancing effect of Zr(IV) on the extraction of Tc(VII) was investigated as functions of the concentration of Zr(IV) in the feed solution, i.e. 0.051.0g•l^-1 and the flow rate of the Tc(VII) stripping solution. A sufficient flow rate of a high-acid strip, i.e.
5 mol•l^-1 HNO₃ is necessary depending on the Zr(IV) concentration to achieve a complete stripping of Tc into an aqueous waste stream. Under a relatively high concentration of Zr(IV) and the high-acid strip, the concentration profile of Tc(VII) showed an accumulation in the process.
To avoid, if necessary, the accumulation of Tc(VII) and a high aqueous flow rate at the scrub part, a by-pass flowsheet, in which the Tc-stripped solution is passed over the scrub part and connected to the extraction stage, was proposed. In the actual reprocessing of spent fuel, determination of the flow rate of high-acid strip and subsequent decision regarding the by-pass mode or non-by-pass mode should be done depending on the amount of Zr(IV) in the fuel to be processed.

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Heat-generating nuclides Cs and Sr were separated from simulated high-level liquid wastes
(HLLW) by successive adsorption on columns of ferrierites (F) and zeolite A. Adsorbed Cs and Sr were efficiently eluted with NH₄NO₃ and EDTA solutions, respectively, yielding the recovery over 96%. A simulated waste solution containing 29 components was denitrated with a formic acid up to pH 7.92. The amounts of nuclides adsorbed from this denitrated solution were experimentally estimated to be 0.33 mmol Cs/g•F and 0.19 mmol Sr/g•A, respectively.

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The control rod worths of the AVR power plant for the cold and hot conditions were analyzed to verify an analysis method of reflector control rod worths. In the analysis method, the neutron-flux-weighting method was used to obtain the effective group constants of the control rod inserted in the graphite nose. The control rod worths were calculated with the core analysis code system for the High Temperature engineering Test Reactor (HTTR).
The experimental values of the control rod worths were 6.47 and 6.81%.4 for the cold and hot conditions, respectively. The differences between the experimental and analytical values for the cold and hot conditions were 2 and 10%, respectively. From these results, it was made clear that the analysis method applied here predicts well the control rod worth of the AVR power plant and is applicable for the nuclear designs of reflector control rods of future small HTGRs.

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An experimental study was carried out to measure large negative reactivities such as
-10%Δk/k using a modified digital reactivity meter. Conventionally such large negative reactivities have been measured by the multi-channel scaler (MCS) method.
In the present study both of the methods were used and the two results agreed quite well.
Thus our new method of utilizing a digital reactivity meter for the measurement has been proved to be useful.

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A slight change in the level-volume relation for an accountability tank for a large amount of plutonium nitrate solution (PuN) was observed at the Plutonium Conversion Development
Facility (PCDF) in the Power Reactor and Nuclear Fuel Development Corp. (PNC), Tokai
Works. From the results of annual tank re-calibrations for the plutonium receiving tank from 1985 to 1992 using the incremental feed of nitric acid as the density standard, it became clear that the relation between the level and the volume changed slightly, and the rate of the change was a linear function of operating time. Also it became clear that the change was linear in relation to the level. In the PCDF, the cumulative change in the volume at the nominal level was evaluated to be 0.1% during 8 years' operation. It was also evaluated that the repeatability of the re-calibration is much better than 0.1%. A reasonable frequency of tank re-calibration is once every 5 years.

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