An evaluation method for the radionuclide content in low level drum package waste by using scattered γ-rays has been investigated. Gamma-rays are counted by a detector after passing through a collimator having appropriate vertical and horizontal openings both for the segmented scan and for the average geometrical efficiency being almost independent of the source position. The attenuation of unscattered γ-rays is estimated from the intensity ratio of scattered rays to unscattered ones of the nuclide emitting the highest γ-ray energy in the drum. Satisfactory results were obtained in numerical simulations and basic experiments carried out using test equipment with a pure Ge detector and a rectangular collimator, and sample 200 l drums, containing sealed ⁶⁰Co and ¹³⁷Cs sources, and having a uniform density of 2.2 g/cm³. The error of the evaluated content, which depended slightly on the activity distribution, was a maximum of about ±30%.

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A fault diagnosis system was developed to improve the availability and maintainability of control systems in nuclear power plants. To facilitate man-machine communications in the system a natural language interface was introduced. Features of the interface include two-step analysis of the meaning of input sentences, identification of the kind of input content using sentence patterns, and retrieval of the information required. For the information retrieval, when information is lacking supplementation is done by referring to the dialog history and knowledge of the control system composition.
Dialog tests, including the required information inputs for diagnosis and questions about the validity of the diagnosis results, were performed by the control system simulator for an assumed component failure, such as a printed circuit board failure. Dialog inputs were made by voice and outputs were by voice or a CRT display. The tests confirmed that inputs required in the diagnosis process can be entered by voice without a conscious limitation of the input sentence style, and that adequate answers can be obtainned in about 1 s allowing maintenance personnel to have a dialog naturally without being aware of any waiting time.

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To evaluate the heat removal capability of an external water wall type containment vessel, which is one of the passive systems for containment cooling, thermal hydraulic behavior in the suppression and outer pools has been examined experimentally and analytically, and the following results were obtained.
(1) The thermal stratification boundary, which separates the pools into an upper high temperature region and a lower low temperature region, was initially observed just below the vent outlet, but later moved downward due to vertical heat conduction.
(2) The effectiveness of a baffle plate, a countermeasure for mitigation of the thermal stratification, was confirmed and its optimal configuration was obtained in the static head-balance model.
(3) The natural convection heat transfer coefficients for downward and upward flow which appear inside and outside of the primary containment vessel wall were measured. These values can be conservatively predicted by the average of Fujii's correlation.
(4) The capability for decay heat removal in the external water wall type containment vessel for a 600 MWe plant was evaluated based on these results, and was found to be large enough.

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The developed high temperature electrical conductivity monitor is applied to the following three items:
(1) Determination of the dissociation constant of an aqueous impurity at elevated temperatures up to 300°C ;
(2) On-line chemical analysis of an aqueous impurity by using the known, significant difference in temperature dependence of electrical conductivity for different chemical forms ;
(3) More reliable prediction of the effects of an aqueous impurity on the corrosion rate at elevated temperatures by using the conductivity at the corroding temperature instead of the one at room temperature.
Obtained results suggest that high temperature conductivity monitoring is effective for implementing an advanced in-service anomaly detection system, using water chemistry data, in nuclear power plants.

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A thermodynamic model has been described for predicting the equilibrium distributions of solute elements in pyrochemical separation systems of molten salts and liquid metals. The model assumes that the solutes are present in the form of complex compounds in the salt phase and of intermetallic compounds in the metal phase. The thermodynamic data of these compounds are available in the literatures and also can be estimated theoretically. Some predictions have been made by the present model to discuss the feasibility of liquid metal extraction system for the recovery of platinum group elements from spent fuel.

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In connection with studies for evaluating the leachability of Cs from solidified radioactive waste, the surface alteration of pollucite was studied under hydrothermal conditions. Pollucite exposed to NaCl or KCl at 300°C formed on its surface a precipitate phase of analcime in the case of NaCl or leucite in the case of KCl. Cubo-octahedral analcime built up not only on the outermost surface of sample but also in the altered layer underneath. This crystal penetration was attributed to precipitation and isomorphous substitution of Na by Cs in the pollucite. The leachability of Cs proved to increase in keeping with salt concentration.
Exposure to CaCl₂ or to MgCl₂ formed a thick precipitate layer of anorthite in the case of CaCl₂ or clinochlore in the case of MgCl₂ over a thin reaction layer. This precipitate layer had the effect of inhibiting the migration of Cs, Si and Al to restrain Cs leaching into solution, which caused the amount of Cs leached to level off beyond 0.01 mol•dm^-3 salt concentration in the case of CaCl₂, and beyond 0.1 mol•dm^-3 in the case of MgCl₂.

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Gamma-ray production data of N, Na, Al, Si, Ca, Ti, Fe, Ni and Cu in JENDL-3 have been tested by comparing the calculated γ-ray spectra with the experimental ones measured at ORNL. As to γ-rays arising from thermal neutron capture, structures observed in the measured spectrum could not be reproduced for most cases with the test version of JENDL-3 (JENDL-3T) whose evaluation was made mainly on the basis of theoretical calculation, while they were reproduced well with ENDF/B-IV. As to γ-rays arising from fast neutron reactions, on the other hand, JENDL-3T reproduced the measured spectra fairly well except for Ca, Fe and Ti, where JENDL-3T gave too high values in the γ-ray spectra between 4 and 6 MeV. Considering the results of the benchmark tests, the γ-ray spectra arising from the thermal neutron capture were reevaluated and the modified data were adopted in the final version of JENDL-3.

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Neutron nuclear data for 15 minor nuclides (Z>88) have been evaluated in the energy range of 10^-5 eV20 MeV. Since only few experimental data are available, the present evaluation was mainly based on the systematics of the data from neighboring nuclides and also optical and statistical model calculations. The evaluations have been carried out for neutron cross sections of total, elastic scattering, inelastic scattering, (n, 2n), (n, 3n), (n, 4n), fission and capture reactions. In addition, angular and energy distributions of the emitted neutrons and average number of the emitted neutrons per fission were also evaluated. The results were compiled in the ENDF/B-V format and stored in the JENDL-3.

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High power CO₂ laser beam was used to simulate high heat flux on a plasma disruption. Fatigue strength and fracture toughness were examined in order to estimate the extent of degradation due to the melted layer, which was the region melted and resolidified in the course of transient high heat flux. The materials were austenitic stainless steels SUS316 and SUS304. Fatigue strength of the irradiated specimen was notably reduced because of the macrodefects on and just below the top surface of the melted layer. However, fracture toughness of the irradiated specimen was not reduced at all. It is considered that melting and resolidification in SUS316 and SUS304 due to high power CO₂ laser beam did not necessarily bring about the degradation of their material strength in the internal part of the melted layer.

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The NEACRP Criticality Working Group has examined the validity of computational methods used in the evaluation of the criticality safety for several different operations involving fissile material through various benchmark problems, and presented a preliminary report summarizing the results from various contributors⁽¹⁾. A hypothetical criticality problem among a variety of benchmark problems proposed by the Working Group represents a system where pellets are surrounded by a fuel solution. A feature of the problem is that the same resonant nuclide exists in adjacent two materials. For methods using multigroup cross sections, this pellet/fuel-solution type system presents a difficulty for treating the resonance absorption in one material which is affected by the adjacent material. The Dancoff factor method is not considered to be valid for the pellet/fuel-solution type system, which evaluates the resonance absorption with the table-look-up method of self-shielding factor based on the narrow resonance approximation (NRA) using the background cross section of the resonant nuclide whose heterogeneity term is estimated by the Dancoff correction factor. Because this method estimates the heterogeneity term of the background cross section by regarding each material as an extreme ; "white" or "black". In the present study, we examine the validity of the Dancoff factor method and two other methods, Tone's method⁽²⁾ and the PEACO method⁽³⁾, for the pellet/fuel-solution type system. A practical method using multigroup cross section was proposed by Tone for a system where the same nuclide exists in different materials. In this method, the background cross section of the resonant nuclide in one material is evaluated by considering a contribution from each material via collision probabilities, that is to say, materials can be regarded as "gray" and the resonance absorption is evaluated by the table-look-up method based on NRA. Tone's method has been reported to give an accurate evaluation for the resonance absorption in the analysis of the critical experiment using fuel elements made of a combination of different kinds of plate-type fuels⁽⁴⁾. The PEACO method implemented in the SRAC code evaluates the resonance absorption by calculating the hyperfine group flux distribution with the collision probability method. The PEACO method has been demonstrated to show good agreement for a large number of benchmarks and reactor calculations⁽⁵⁾. The validity of these methods for the pellet/fuel-solution system is examined in comparison with a continuous energy Monte Carlo code VIM⁽⁶⁾ used as the reference calculation.

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To examine the performance of the high efficiency particulate air (HEPA) filter, demonstration tests were done in various conditions, such as ; normal condition, high temperature and earthquake condition, using conventional deep-pleat filters, 610-×610-×292-mm in size.
The authors found that the HEPA filter acted efficiently enough in various conditions supposed in the air purification system of a nuclear plant. Some empirical formulae were proposed to express the performance of the filter.
The test results were summarized from these three points : (1) decontamination factor (DF) of intact filters, (2) mechanical strength and (3) pinholes, because these three points are important criteria in evaluating the performance of HEPA filter systems.

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