A new nodal SN transport method has been developed so as to perform accurate transport calculations in three-dimensional (3-D) hexagonal geometry. In the present method the neutron angular distributions of intra-node fluxes and transverse-leakage are represented using the S_N quadrature set, and the spatial distribution of neutron source is approximated by a quadratic polynomial expansion. Additionally, the authors have applied the nodal-equivalent finite difference algorithm to 3-D hexagonal geometry in order to establish a stable and efficient iterative scheme,
The present method has been applied to a fast reactor benchmark problem. The results of the present method agree well with those of a Monte Carlo method, the difference in k_eff being less than 0.05%Δk. This shows high accuracy of the present method.

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The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a water-filled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the ANISN, DOT3, 5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses.

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4πβ-γ coincidence counting using a 4πβ pressurized proportional counter combined with the computer discrimination method has made it possible to determine the disintegration rate precisely and rapidly. Recently, measurement of short half-life nuclide requires further shortening of the gas replacement time, and we proposed to use atmospheric CF₄ gas instead of pressurized Ar 90%-CH₄ 10% mixture. However, a conventional 4πβ counter with CF₄ gas at atmospheric pressure gave unsatisfactory results ; small slopes of coincidence efficiency functions and low detection efficiencies for electron-capture decay nuclides. In order to resolve these problems, we constructed a new 4πβ counter with large volume and multianodes. Good plateau characteristics and good time relation between β and γ pulses were obtained in spite of large volume, and the slopes of coincidence efficiency functions and the detection efficiencies were improved. Use of this counter made it possible to measure short half-life nuclides and also easy to measure usual nuclides.

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A unified theory is presented for diagnosis of power reactors with feedback mechanism. In terms of correlation functions calculated from stationary time series data, a feedback system can be expressed by an equivalent innovation model. From the formalism the identifiability of open loop transfer functions in the feedback system is discussed under conditions ; (1) the minimum phase system, (2) the independence of noise sources. From the feedback structure, the property of the nonminimum phase is related to that the reactivity is positive or zero-power reactor transfer function is supercritical. In the case of the nonminimum phase and correlated noise sources, there is a possibility of new additional loops in an innovation model equivalent to the feedback system.

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Highly-dense spherical particles of thorium-based oxides, ThO₂ and (Th, U)O₂, prepared by the sol-gel method were subjected to dissolution with nitric acid containing 00.05 mol/l NaF at high temperatures above 120°C. The dissolution rate depended upon temperature, fluoride concentration and UO₂ content. High-temperature in the range of 120200°C enhanced the dissolution of the ThO₂-based fuels. At low temperatures and/or low UO₂ concentrations, insoluble tetrafluoride precipitates were formed on the particle surfaces and they resulted in the decrease of the dissolution rates. In the present study, the apparent activation energies for the high-temperature dissolution were obtained.

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A theoretical study on the electronic structure of zirconium oxide using a molecular orbital method was carried out to investigate the additive element effects on the electronic conductivity of oxide film formed on Zr-alloys. The atomic clusters used were (MZr₁₂O₈)³⁶⁺ (M=Zr, 3d-transition metals and alkali metals). To simulate the electron conduction process in the oxide, calculations for a cluster with oxygen vacancy (Vo) were also carried out. The energy gap E_g between electron-occupied and empty levels was evaluated, and the electronic conductivity was estimated qualitatively. Opposite effects on the electronic conductivity were found for additions of 3d-transition metals and alkali metals. The latter increased the electronic conductivity by forming impurity levels with small E_g. The former, however, induced compressive strain in the oxide, resulting in a lowering of electronic conductivity due to widening of the energy gap at the oxygen vacancy.

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An effective pre-oxidation method for Alloy X750 was developed to reduce general corrosion in an oxygenated aqueous environment such as in BWR core water. The optimum condition of preoxidation in air at elevated temperatures was found to be 520 h at 973 K by considering the allowance condition of heat treatment for age-hardening.
Some characteristics of the corroded oxide film have been clarified by surface analyses with XMA, SIMS, AES, XPS etc. The film was composed of double oxide layers, namely a highly crystallized NiFe₂O₄ outer layer and a high Cr₂O₃ content inner layer. The passive property of the film has been recognized to be due to the nature of the oxides whereby NiFe₂O₄ restricts the dissolution of metals because of its low solubility and Cr₂O₃ restricts the diffusion of metal ions because of its high binding energy and low diffusion coefficient.

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A simulation model was proposed to study the mass and heat transfer rates in hydrogen isotope distillation columns. The model is on the basis of the analogy : the mass transfer coefficients and effective surface area within the column were estimated by Kosuge's and Onda's correlation in the model. The calculated results by the present model agreed approximately with the experimental observations. The model had a tendency to give slightly higher separation performance. This tendency is due to that the estimated effective surface area might be larger than actual values. The calculated results by the proposed model were compared with those by the stage model also to discuss relationship between the mass transfer rate and the HETP value. The estimated mass transfer rates were almost the same for H-D and D-T systems. This calculated results give a reason why the HETP values were not varied for the difference in physical properties of hydrogen isotopes. The proposed model does not have the parameter that fit the calculated results to the experimental observations. Comparing the calculated results by the proposed model with those by the stage model, we can estimate the HETP value.

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When uranium vapor is generated with an electron beam evaporator, a uranium plasma is formed on the evaporating surface. This plasma rises and expands with the vapor. Propagation behavior of this plasma was investigated by measuring plasma parameters, drift energy of ions and vapor flux along the propagation path. Over the range of 20-50 cm from the evaporation surface, the plasma density decreased from 3 × 10⁹ cm^-3 to 3 × 10⁸ cm^-3, while the electron temperature had a constant value of 0.29 eV. When the space potential was lowered from 1.48 to 0.80 V, the plasma ions were accelerated to increase the drift energy from 1.50 to 2.14 eV. Validity of the Boltzmann electron distribution was checked by comparing the space potential distribution with the plasma density distribution, and also the floating potential distribution with the ion flux distribution. These results confirm that the ambipolar diffusion governs the plasma propagation behavior. The change in the plasma density during its propagation occurred not only by an increase of plasma volume, but by the ion acceleration toward the propagation direction as well.

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The fate of radionuclides produced in a light water moderator was investigated by assaying with a Ge(Li) detector sampled coolant water and the waste water produced by regeneration of resins, measuring the radiation dose rates along the surface of the resin column length with TLDs, and studying the sorption of radionuclides to resins in dynamic experiments. The buildup concentrations of short-lived radionuclides, produced by activation in the reactor core water and the amounts of radionuclides sorbed on the resin bed, as estimated from a compartment model, are reported. A method for estimating the production rate of soluble radionuclides with long half-lives in the core water has been developed. Changes in the radiation dose rates along the resin column length recorded during reactor operation and after shut-down, showed that the sorbed radionuclides did not migrate downward, as expected from the chromatography theory, rather they were decreased by the decay process. Moreover, the slopes of the dose rates on both sides of the resin surface were almost identical. These results show the behavior and distribution of radionuclides in the light water moderator of a research reactor.

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The paper deals with the activity acquired by an element of target area moving parallel to a flat disk-shaped source isotropically emitting in nondispersive media. Some particular cases of motion are theoretically investigated in order to establish analytically the activity acquired in given irradiation periods.

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For the precise calculation of the burnup of minor actinide isotopes, a code system-SWAT has been developed. This system analyzes burnup problems with neutron spectrum that depends on the type of a reactor and the irradiation history, using latest evaluated nuclear data files JENDL-3 or ENDF/B-VI. The post irradiation test in TRINO and the recent experiment in typical PWRs in Japan were analyzed with SWAT. These analyses show that the results of U and Pu for high burnup fuels almost agree with experimental results but those for middle burnup fuels do not agree with them. The results for Am and Cm isotopes still have large discrepancy. The average C/E of ²⁴³Am is 0.79, and that of ²⁴⁴Cm is 0.70 for high burnup (33, 000 MWd/tU) samples.
For middle burnup (25, 000 MWd/tU) samples, the C/E for ²⁴⁴Cm is over 2.0. The discrepancy is partially explained by considering the power peaking history of first cycle and second cycle.

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The Pt-Mo alloy thermocouples with element wires of Pt-5%Mo and Pt-0.1%Mo were developed for incore temperature measurements at a maximum temperature of 1, 200°C for a long term and of 1, 350°C for a short term in VHTRs. The thermocouples were tested at high temperature conditions of 1, 0001, 400°C in an outpile electric furnace and in an incore capsule in the JMTR. The total thermal neutron fluence irradiated was 8.1×10²⁰ n/cm². The changes in their electromotive forces due to those exposures to high temperatures and radiation in the reactor core were evaluated to be less than ±2%. This meets the specifications that is required for the incore temperature measurements in VHTRs. Hence, the developed thermocouples can be used also in the incore temperature measurement in advanced type HTGRs and in incore irradiation experiments in the HTTR, in future.

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