The β-ray spectra of individual fission products were calculated by using the β-decay data assuming every β-decay to be allowed transition. For the nuclides without measured decay data the β-feeding function was evaluated with the gross theory of β-decay and the β-ray spectrum was calculated from the function. The measured decay data were also supplemented with the data calculated by the gross theory for the excitation energy range above the highest measured excitation energy level. The β-ray spectra from aggregate fission products after a burst fission were calculated by using the β-ray spectrum and the atom number of each fission product nuclide and they were compared with the ones measured for thermal neutron induced fission of ²³⁵U, ²³⁹Pu and ²⁴¹Pu at Oak Ridge National Laboratory. The spectrum calculations showed excellent agreement with the measured data at shorter cooling times than 10 s when many short-lived nuclides without measured decay data contributed considerably to the spectrum.

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A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance.
This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs.

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The vectorization method was studied to achieve a high efficiency for the precise physics model used in the continuous energy Monte Carlo method. The collision analysis task was reconstructed on the basis of the event based algorithm, and the stack-driven zone-selection method was applied to the vectorization of random walk simulation. These methods were installed into the vectorized continuous energy MVP code for general purpose uses. Performance of the present method was evaluated by comparison with conventional scalar codes VIM and MCNP for two typical problems. The MVP code achieved a vectorization ratio of more than 95% and a computation speed faster by a factor of 822 on the FACOM VP-2600 vector supercomputer compared with the conventional scalar codes.

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Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a threedimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio).
The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit.
A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide-oscillation, because of the sensitivity difference in a flow-to-power gain.
A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals.

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Solubility of sodium iodide in liquid sodium was determined over a temperature range of 350800°C by using equilibration and sampling techniques.
The dissolution of a 23 g fragment of sodium iodide crystal into 7.58 g sodium was made in a base cell of reversed "J" type-stainless steel capsule, which was heated in an electric resistance furnace. After the equilibration had been established, a portion of the solution was decanted into a side arm cell of the capsule by tilting the capsule about 110° with the furnace capable of rotating round its horizontal axis. The sodium transferred to the arm cell was analyzed as a single sample.
The solubility S in the unit of ppm by weight is expressed as a function of temperature in the unit of Kelvin in the forms, log₁₀S=8.46-3, 440/T above 660°C,
log₁₀S=9.65-4, 550/T below 660°C,
which fit well with Bredig's data in the higher temperature region. The partial molar excess quantities, calculated from the thermodynamic properties of sodium iodide and its individual solubility data, are in reasonable agreement with those determined for the dilute solution of the iodide in liquid sodium by Castleman & Tang.

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Calibration curves of extremely low concentrations of the alloying elements Sn, Fe, Cr and Ni in Zircaloy were obtained, using standard samples, by energy dispersive X-ray spectroscopy to measure concentration distributions of alloying elements dissolved in the Zircaloy matrix, Their detectable limits were 0.21 at% for Sn, 0.06 at% for Fe, 0.04 at% for Cr and 0.03 at% for Ni. Then concentration distributions of alloying elements in unirradiated and neutron irradiated Zircaloy-2 were measured using these calibration curves. It was confirmed that neutron irradiation increased the dissolved concentrations of Fe, Cr and Ni. Furthermore, Cr diffused slower than Fe and Ni. It was suggested that the rate limiting process of irradiation-induced dissolution from Fe, Cr-type precipitates into the matrix was the diffusion of alloying atoms in the precipitates and that the dissolution process proceeded due to displacement of alloying atoms from the precipitates into the matrix and diffusion in the matrix.

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Carburization and decarburization behavior in sodium have been examined for Fe-9Cr-Mo ferritic steels exposure to flowing sodium with containing 0.0049, 0.016 and O. 16 wppm carbon at temperatures between 723 and 873 K.
The Fe-9Cr-Mo ferritic steels were more resistant to carbon transfer than 21/4Cr-1Mo steel. The carbon transfer and its gradient in Fe-9Cr-Mo ferritic steels was closely related with carbon concentration in sodium, and carburization and decarburization behavior depended on the sodium conditions. In the case of the application of Mod. 9Cr-lMo steel to heat transfer tubes of steam generator of LMFBR, carburization behavior was observed on the conditions that sodium inlet temperature was 778 K and carbon concentration in sodium was 0.010. 10 wppm. The mean carbon contents of heat transfer tubes after the operation of the steam generator for 200, 000 h were estimated to be about O. 1030. 148 wt% comparing with initial carbon contents of 0. 10 wt%. Consequently it is expected that high-temperature strength of Mod. 9Cr-1Mo steel would retain sufficient strength and ductility for such operating conditions.

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The feasibility of remote, in-service monitoring of as-machined nuclear graphite parts using noncontacting laser ultrasonics was investigated. The anisotropy sensitivities of laser ultrasonics and several other conventional techniques were compared for three significantly different nucleargrade graphites. Laser ultrasonics using a Fabry-Perot interferometer proved sufficiently sensitive to detect the anisotropy in a bulk graphite having moderate grain size. In contrast, a coarsegrained graphite was difficult or impossible to investigate using any of the techniques. Laser ultrasonic monitoring of some graphites may be useful in future nuclear operations, particularly for in-service applications where contacting and destructive analyses are unsuitable.

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To improve the performance of the roof slab for LMFBRs, information on the mass concentration and particle size distribution in the gas space enclosed by the reactor vessel is important, however experimental data are very limited⁽¹⁾⁽²⁾. Himeno⁽¹⁾ measured the mass concentration and particle size in the gas phase using small sodium pots. Schi_itz et al.⁽²⁾ also tried to determine the mass concentration and the particle size distribution at high temperatures. However, mist behavior in the cover gas is very complicated and those values in the actual FBR cover gas have not been completely clarified. A few efforts have been devoted to solving the equations that govern particle size distribution of the mist. Himeno⁽¹⁾ and Kudo⁽³⁾ obtained a solution by representing the mist size with its geometric diameter. However, in this method, the governing equations become very complicated and a large numerical error may be induced because the addition/subtraction law does not hold for the change in the geometric diameter due to each coagulation.

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In order to simplify the disposal procedure of high level radioactive waste produced by spent fuel reprocessing, it is necessary to separate trans-uranium elements (TRH) from high level liquid waste (HLLW). Neptunium (Np) (V) is the dominant Np oxidation state in HLLW. Since extractability is low, its separation is extremely difficult. However, valence control of Np(V) to (IV) or (VI) allows Np separation. In a previous study on reduction kinetics of Np(V) to (IV) by hydroxylammonium in nitric acid solution, the activation energy was reported to be 105 kJ/mol⁽¹⁾. This value suggested that the rate of Np(V) reduction was very slow at room temperature. Although, the coexistence of platinum black with hydroxylammonium nitrate was found to give a remarkable increase in the reduction rate of Np(V) in nitric acid solution⁽²⁾.
In the present study, the kinetics and mechanism of the reduction reaction by platinum (Pt) black catalyst are studied, and the rate constant at room temperature is determined.

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For the estimation of external photon exposure, it is important to assess the dose equivalent in organs or tissues H_T and the effective dose equivalent H_E while the accuracy required will depend on the use of the data⁽¹⁾. As the quantities of the dose limits are not measured directly, conversion coefficients that link dosimetric and radiation protection quantities are necessary to estimate them. Two approaches to determine the conversion coefficients, computational and experimental, have mainly been used, but an evaluation of these is not a trivial matter.

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This paper describes an automatic flaw detection equipment and heat-resistant ultrasonic transducer for plate thickness measurement. The automatic flaw detection equipment is used during in-service inspection. It comprises an angle-beam electromagnetic acoustic transducer (EMAT), mounted on a vehicle. for scanning the pipe surface to be inspected. The EMAT functions without direct contact with the pipe surface through a coupling liquid, the vehicle does not require a guide track installed on the pipe surface, since it is equipped with magnetic wheels that adhere to the pipe, permitting it to travel along the circumferential weld joint of a carbon steel pipe. Another heat-resistant ultrasonic transducer is a normal beam EMAT and is used during plant operation, As a result, the automatic flaw detection equipment could detect a 1 mm deep notch cut on a test piece of 25 mm thick carbon steel plate. The vehicle location accuracy on the piping was ±2 mm. The normal beam EMAT could measure the plate thickness, within ±0.3 mm accuracy for the range of plate thickness 4 mm to 12 mm at 300°C.

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