A series of benchmark tests was made to check the neutron nuclear data of main fissile nuclides (²³⁹Pu, ²³⁶U and ²³³U) of JENDL-3 for fast reactors. A total of nine critical assemblies were analyzed. They are assemblies of single material, high enrichment and simple geometry with small volume and therefore suitable for nuclear data testing. Criticality calculation was made by ANISN with S₁₆P₅ using the VITAMIN-J 175-energy-group. Discussions are made on ken, spectral indices at core center and leakage spectra.
From the study, a problem was pointed out relating to the interpolation of secondary-neutron energy distributions for threshold reactions near the threshold energy point adopted in the original JENDL-3 and its remedy was proposed. By the benchmark tests of thus revised JENDL-3 (JENDL-3.1), it was shown that integral experiments for ²³⁹Pu and ²³⁵U cores were reproduced quite satisfactorily. On the contrary, it was revealed that large deviations for ²³³U cores from the experiment were due to uncertainties of the fission spectrum and the inelastic scattering cross sections, In the present work, sensitivity of "a" parameter (level density parameter) of Madland-Nix's fission spectrum formula to the integral data was extensively studied. Some recommendations are made to improve JENDL-3.1.

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To obtain fundamental data for the research of the transmutation of nuclear waste, the thermal neutron cross section and the resonance integral of the reaction ¹³⁷Cs(n, γ)¹³⁸Cs have been measured by means of a modified Cd-ratio technique. Targets of about 0.4 MBq ¹³⁷Cs were irradiated with neutrons in the JRR-4 swimming pool reactor by using a pneumatic tube equipped with a movable Cd shield, and neutron fluxes and their epithermal neutron fractions were monitored with activation detectors (Co/AI and Au/AI) whose sensitivity to epithermal neutrons differed from each other. After irradiation, the ¹³⁷Cs samples were purified chemically and their γ-ray spectra were measured with a high purity Ge detector. The resulting yields of ¹³⁸Cs for different neutron spectra and the neutron flux data have yielded that the thermal neutron cross section (for 2, 200 m/s neutrons) is 0.25±0.02 b and the resonance integral 0.36± 0.07 b. The results are consistent with the effective cross section (σ=0.250±0.013 b) of our previous measurement. An uncertainty due to neglecting the production of the short-lived isomer of ¹³⁸Cs was calculated on the basis of the isomeric yield ratio given by the Huizenga and Vandenbosch model. A probable uncertainty was estimated to be +1.2% or +3.9%, depending on the spin of the compound nucleus produced in the s-wave neutron capture.

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This paper presents a new approach using the bifurcation theory for the stability analysis of BWRs. In this approach, the dependencies of the equilibrium states on the parameters that have a large influence on the stability are investigated topologically over a wide range of phase space. The stability information can be derived from the analysis of the bifurcation phenomena on the equilibrium states. This investigation enabled us to obtain qualitative and global information on the stability of a nonlinear system. The new approach was applied to the analysis of the stability associated with in-phase power oscillation (core reactivity stability). The loss of linear stability took place at a lower reactor power as the coolant flow rate decreased, and this instability occurs at the Hopf bifurcation point. The sensitivity analysis of the stability boundary for the various parameters revealed that the channel hydrodynamics heavily play a significant role in the stability. The Hopf bifurcation analysis proved that the periodic state bifurcating at the Hopf bifurcation point was orbitally unstable and a limit cycle attractor did not exist in the vicinity of the bifurcation point. This fact led to the conclusion that a limit cycle in-phase power oscillation observed in BWR instability was not excited directly by the bifurcation of the orbitally unstable periodic state.

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This paper deals with the results of critical power experiments performed simulating commercial BWR conditions at Hitachi Core and Fuel Test Loop with seven fuel-rod heater bundles in a tight triangular lattice array, which is assumed to be used in high conversion BWRs. The main conclusions of experiments are;
(1) The critical power in high conversion BWR fuel bundles in a tight triangular lattice array showed almost the same dependence on operating conditions, namely mass flux, pressure, inlet subcooling etc., and the critical power can be accurately correlated by critical quality and boiling length relation as in current BWR fuel bundles in a loose square lattice array,
(2) A correlation to predict the critical power in high conversion BWR fuel bundles was obtained from the experimental results, and the predicted results by the correlation were summarized as,

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Laser induced photoacoustic spectroscopy was applied to the determination of ⁹⁹Tc in a solution coupled with quantitative complexation of Tc with to (thiourea) to form Tc(tu)₆³⁺.The complexation procedure was adopted to shift the uv absorption peak of pertechnetate to the visible region. Photoacoustic signals due to the 488 nm absorption peak of Tc(tu)₆³⁺ was analyzed by using a Fast Fourrier Transform (FFT) system. The quantitative detection of Tc could be easily performed down to the 10^-8 M concentration in this system. This is about 1, 000 times more sensitive than the ordinary colorimetric measurement.

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The migration of polystyrene latex particles (diameter=310, 170 and 40 nm) through columns packed with 30 μm quartz powder was experimentally studied at low eluant velocity at pH 6.0±0.1. The average velocity of polystyrene latex particle was found to be larger than that of HTO. The ratio (separation factor) of the average velocity of polystyrene latex particle to that of HTO increased with increasing the size of polystyrene latex particle and with decreasing the concentration of sodium chloride added under the experimental conditions. This reveals that the colloid migration process in the geological media is similar to that occurring in hydrodynamic chromatography. A model proposed for hydrodynamic chromatography was used to calculate the separation factor in which the potentials from tile double layer force and from the van der Waals force were taken into account. The separation factor, calculated with the model by using the measured surface potentials of polystyrene latex particle and quartz powder and by using Hamaker constant evaluated from the literature, was found to agree well with the experimental result.

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A sintered product of cesium aluminum silicate, CsAlSi₅O₁₂, formed an outer precipitated layer and a reaction layer when exposed to 3.0 ti (=mol/dm³) NaCI under hydrothermal conditions above 250°C to 300°C. The precipitated layer consisted of crystallized phases of analcime and montmorillonite, and the reaction layer exhibited a concentration gradient for Cs. These layers grew with temperature and reaction time. When the solid sample was contacted with NaCl solutions below 0.1 M, the crystallized phase of pollucite was formed in the precipitated layer. Successive accumulation of pollucite layer contributed to the depression of leaching of Cs.

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The sorption characteristics of ²³⁷Np by sodium-smectite have been studied by batch type sorption-desorption experiments over a pH range of 2 to 8. The association of ²³⁷Np with smectite was examined by the sequential extraction technique in which 1 m KCl and 1 Nit HCl solutions were used as the extraction agents. Two kinds of neptunium-smectite association were identified by their characteristic desorption properties. In the first association, neptunium is desorbed from smectite by 1 m KC1. In the second association, neptunium is stable in the basic solution but is desorbed by 1 Ni HC1. The latter association was observed when neptunium was sorbed at low pH. Powder X-ray diffraction showed an expansion of the basal spacing of smectite from 1.25 to 1.48 nm in low pH solutions, consistent with specific sorption of neptunium at a H⁺ expanded interlayer of smectite.

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By using the two-dimensional rigorous numerical solution of equations of change, the change in hot-wire temperature caused by flow circulation near the top and bottom plates of thermal diffusion column with 14.7 mm-radius and 80 K cold-wall was analyzed with constant heat flux condition on hot-wire temperature boundary by parametrically changing the pressure. Flow field analysis for F1, -HD (tracer level) gas mixture in total-reflux operation has revealed : ( i ) The magnitude of the change in hot-wire temperature is almost proportional to the square of the pressure, while the magnitude of the free convection is almost proportional to the pressure. (ii) The magnitude of the change in hot-wire temperature at the bottom is larger than that near the top plate, while the magnitude of the radial flow component is almost the same at the top and bottom, and (iii) The range of temperature change is wider than the range at the top and bottom part of the column where convective flow turns around (i.e. radial velocity components significantly exist).

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In order to accurately calculate effective neutron cross sections in the resonance energy region, the multiband method has been applied to cell calculations. Cell calculations for UO₂ and MOX fuels of light water reactors have been performed and the results were compared with those of a continuous energy Monte Carlo code VIM and the conventional self-shielding method using the Dancoff factor.
The k_∞ values calculated by the multiband method agreed with those of the VIM calculations within 0.20%Δk for the UO₂ fuel cell and within 0.30%Δk for the MOX fuel cell, respectively, whereas the Dancoff factor method yielded about 1.1%Δk errors for the two cells. The elementwise contribution to this error was investigated, and it was found that the effective microscopic cross sections, particularly those for the giant resonances of ²³⁸U, calculated by the multiband method were in good agreement with those of VIM. It was also found that interference effect between ²³⁸U and ²³⁵U resonances in the UO₂ fuel and that between ²³⁸U and ²³⁹Pu resonances in the MOX fuel made about 0.20% Δk contributions to k_∞ in both fuel cells.

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Horizontal extrapolated distances λ_h and critical bucklings B²_C for light-water moderated low enriched UO₂ cores were evaluated using the variable loading method. A set of critical height (water levels) of the cores having various horizontal dimensions was used in this method. Errors of the present method were smaller than the previous ones by the flux shape method in which the horizontal extrapolated distance was determined from the horizontal power distribution. The presently obtained λ_h and B²_C reproduced well the change in the critical water level for the whole range of the horizontal core size, while the previous ones could be used only for a limited range. Moreover, the effective multiplication factors obtained through the cell calculation of the SRAC code by using the presently evaluated critical bucklings agreed much better with the rigorous ones by the continuous energy Monte Carlo calculation using a full core model, than those by using the previous values. Therefore, it was concluded that the variable loadin, g method can give more accurate values for λ_h and B²_C than the flux shape method for the cores investigated.

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Reactivity control of the Fugen Nuclear Power Station is performed by the control rods and the adjustment of the concentration of 10B dissolved in the heavy water as boric acid.
An accurate evaluation of the ¹⁰B concentration in the heavy water is important in order to determine the excess reactivity.
It has been a conventional method to sample the heavy water and measure the ¹⁰B concentration in it directly, which is time- and labor-consuming and, moreover, costly. We have developed an economical and expedient calculation method to determine the ¹⁰B concentrations in the heavy water. This method uses a model of boric acid injection, boric acid removal, poison burning in the core and heavy water purification system.
The calculation result has been evaluated by the comparison of calculated ¹⁰B concentrations with the data obtained by mass spectrometry method, and the calculation accuracy is approximately ±0.1 ppm. This calculation program is practical and successfully in use now.

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High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950°C in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly.

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