The parameterization of the double differential cross section is made for the neutron emission from proton-induced spallation reaction. The neutron data for incident proton energy region of 600800 MeV are well analyzed by the moving source model on the basis of the Maxwell-like energy distribution. In the incident energy region below about 400 MeV, however, the use of this method leads to an unsuccessful parameterization result. Then, the Watt distribution is introduced into the moving source model in this energy region. The combination of the two distributions enables us to parameterize the double differential cross section of neutron emission for the incident protons of 25 to 800 MeV and to find the systematics of moving source model parameters for a wide variety of target nuclei.

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The author examined the validity to estimate the subcriticality of a test region in a coupled reactor system using only measurable quantities on the basis of Avery's coupled reactor theory. For the purpose, we analyzed coupled reactor experiments performed at the Tank-type Critical Assembly in Japan Atomic Energy Research Institute by using two region systems and evaluated the subcriticality of the test region through a numerical study. Coupling coefficients were redefined at the quasi-static state because their definitions by Avery were not clear. With the coupling coefficients obtained by the numerical calculation, the multiplication factor of the test region was evaluated by two formulas; one for the evaluation using only the measurable quantities and the other for the accurate evaluation which contains the terms dropped in the former formula by assuming the unchangeableness for the perturbation induced in a driver region. From the comparison between the results of the evaluations, it was found that the estimation using only the measurable quantities is valid only for the coupled reactor system where the subcriticality of the test region was very small within a few dollars in reactivity. Consequently, it is concluded that the estimation using only the measurable quantities is not applicable to a general coupled reactor system.

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This paper describes an on-line algorithm for predicting the critical control rod pattern, which has been developed to reduce the mental strain on operators while withdrawing control rods in the BWR plant startup operation. The proposed algorithm estimates a target eigenvalue (eigenvalue bias) for a three-dimensional neutron kinetics model with a neutron source incorporating actual neutron detector readings. The critical control rod pattern is then predicted based on the estimated eigenvalue bias. The algorithm has been verified using data obtained from an actual startup operation on a BWR model-5 plant, and the estimated eigenvalue bias agreed well with the effective multiplication factor at the criticality actually determined from the operator's judgement.

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The authors have studied the dynamical features of limit cycle oscillations in a nonlinear reactor system equation with feedback time delay in both control system and temperature effect by taking account of the experience in the pulsed reactor NSRR before its recent modification. The features of competition between the two delayed feedback effects are analyzed. A theoretical method for predicting analytically the mean amplitude of limit cycle oscillation is also developed by combining the method of Krylov-Bogoliubov and that of nonlinear transformation. Further, the effects of colored external noise on the profiles of power oscillations is investigated from the viewpoints of "global stability" and "non-linear response". Their new findings are summarized as follows:
(1) Without an external noise, chaotic oscillations can arise when the product of the applied reactivity and the delay-time is very large. However, it does not appear when this product term takes a small value.
(2) On the other hand, near the first Hopf bifurcation under the influence of external noise, the limit cycle oscillation becomes chaotic one as a superposition of quasi-periodic oscillations, which can be identified by the stochastic Van der Pol equation. The sinusoidal limit cycle oscillation of reactor power observed in the NSRR has such a stochastic nature rather than kinematic one.

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Presented are results of an application of a new primary coolant inventory tracking method which was invented for a Westinghouse-type pressurized water reactor, to various kinds of small break loss-of-coolant accident experiments conducted at the Large-Scale Test Facility in Japan Atomic Energy Research Institute.
The uniqueness of this method is that it can track the primary coolant depletion prior to the initiation of inadequate core cooling. The primary coolant inventory is tracked by measuring the water level in the vertical region of each primary loop including the steam generator outlet plenum and by using a simple correlation between the level and coolant inventory. The principal level measuring range corresponds to the primary coolant volumes of approximately 30 to 60% of the initial volume. A limitation of the reactor vessel level indication system is also shown in comparison with this method.

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The precipitation behavior of Pu, Np and Am during the denitration of high-level radioactive liquid waste (HLW) by formic acid was studied using a simulated HLW. The dissolution of the precipitate formed in denitrated HLW was also studied using oxalic acid to recover transuranium (TRU) elements from the precipitate. In the denitration, the precipitated fractions of TRU elements increased with decreasing acidity of the denitrated HLW. In the denitration at [HCOOH]/[HNO₃]=1.5, which was adopted in the partitioning process developed in JAERI, the precipitated fractions of Np and Am were only 0.6% and 0.06%, respectively, whereas that of Pu was 90%. The precipitation fractions of Pu and Np did not depend on their concentrations in the range of 6×10^-56×10^-4M for Pu and 10^-510^-3M for Np. Plutonium was not precipitated itself by polymerization or hydrolysis but coprecipitated with other elements such as Mo and Zr. It was found that the precipitate formed during the denitration of 1l of HLW could be dissolved in a 800 ml of 0.5 M oxalic acid solution.

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Modeling was carried out on activities of components in the nitric acid-nitrate-water ternary system in order to examine details of the chemical reactions in reprocessing solutions. To modify our earlier "Hydration Model", the ternary system was regarded as the mixed solution of two kinds of cluster molecules which were dominant associated molecules in the nitric acid-water system and nitrate-water system, respectively. Activities in the ternary system were calculated from those in the two binary systems taking into account the change of entropy and enthalpy during mixing. The entropy term was calculated assuming ideal mixing. The enthalpy term was calculated assuming that the nitric acid cluster would be influenced by one nitrate cluster as if there were m_s/m_na numbers of nitric acid clusters, where m_s and m_na were the numbers of water molecules in nitrate and nitric acid cluster molecules, respectively.
The model was able to predict activities of nitric acid, water and nitrate. Its use was verified for the activities of nitric acid and water provided phase separation did not occur. The verification for nitrate activity was not carried out, however, the model should be suitable.

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The effect of temperature on the entrainment of Ru, Tc and Se was studied in calcination experiments performed with continuous feeding of simulated high-level liquid waste containing radioactive tracers. The calcining temperature ranged from 300 to 800°C. The entrainment ratios showed a tendency to peak around 300400°C followed by a dip and a second rise beyond 500°C. The volatilized fraction of Tc and Se increased with rising calcining temperature, whereas Ru volatilization decreased with rising temperature in the range covered by the experiment, i.e. above 300°C.

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1. Introduction
The quantification of the erosion and redeposition processes during plasma-material interaction is a major issue which has to be resolved to provide a design basis for the plasma facing components of future magnetic fusion devices. Values for the plasma erosion of the divertor armour due to ablation by incident plasma ions are of the order of meters/burn year for ITER. This would result in prohibitively short lifetimes for the plasma facing armour. However, it is hoped, that most of the ablated material will be ionized by the incident plasma in front of the divertor plate. These ions would subsequently be guided along the magnetic field back to the material surface and be redeposited there. The repeated redeposition of the eroded material would effectively reduce the net erosion of the plasma facing material. For the analysis of this erosion and redeposition process numerical models of the plasma-material interaction and the impurity transport in the vicinity of the divertor plate have been developed⁽¹⁾. These models indicate that redeposition may reduce the erosion loss on the divertor by a factor of about 10. Until now the validity of the erosion and redeposition model is still questionable, since no quantitative experimental evidence for the redeposition process is available. Qualitative indication of material redeposition has beenfound in tokamaks, e.g.⁽²⁾, and a reduction of the erosion yield under plasma irradiation was reported from PISCES⁽³⁾.
The purpose of the present laboratory experiment was to give quantitative evidence for the redeposition of eroded material under plasma irradiation. The principle was to expose a carbon tracer specimen to a plasma of high density and to separate the locations of preferential erosion and redeposition. Thus the redeposition of the eroded carbon took place on an inclined metallic probe plate and could be quantified subsequently by surface analysis. The data obtained by such experiments can be used for basic studies of the erosion and redeposition of plasma facing materials on one hand, and on the other hand for benchmarking of computer codes.

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The concept of a direct cycle light water reactor (LWR) operating at supercritical pressure is presented. It is attractive for improving the thermal efficiency of LWRs and for simplifying the reactor system⁽¹⁾. The critical pressure of water is 22.1 MPa (221 bars). The density of water changes continuously above it and the concept of boiling does not exist. The system is conceptually depicted in Fig. 1. The water coolant which flows into the core is heated up and can be directly fed to turbines. Compared with the current BWRs, the recirculation system and steam separators and dryers will be eliminated. The vessel size will be greatly reduced. This will reduce the capital cost. The thermal efficiency will be also revolutionally improved. The feasibility of the system was studied through the conceptual design of the core.
The axial temperature and density distribution of the coolant in the core were calculated at 25 MPa (250 bars) using a single channel model of a fuel and coolant. The density decreases continuously from the inlet (0.725 g/cm³) to the outlet (0.137 g/cm³) and the temperature increases from 583 K (310°C) to 689 K (416°C). The heat is efficiently removed due to the high specific heat of water around 658 K (385°C). The density change is mild at 250 bars than 225 bars which is close to the critical pressure. It is desirable to choose 250 bars for the stability of the flow in the core.

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Nodular corrosion on the surface of Zircaloy-2 fuel cladding tubes was investigated and the effects of irradiation time, linear heat rate, and fast neutron flux on the nucleation and growth of nodules were statistically studied. The nucleation and growth rate of nodules were found to decrease with irradiation time. Then, the nodular corrosion resistance of neutron irradiated Zircaloy-2 was examined in an out-of-pile corrosion test, and neutron irradiation was seemed to improve the Zircaloy-2 nodular corrosion resistance. Finally, microstructural evolution of Zircaloy-2 during neutron irradiation was investigated and radiation-induced dissolution of intermetallic precipitates was confirmed. It was proposed that an increase of the dissolved alloying element concentration should improve the nodular corrosion resistance of Zircaloy-2 and restrain the nucleation and growth of nodules on Zircaloy-2 fuel cladding tubes during irradiation in BWRs.

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