A planar hybrid undulator was constructed for the JAERI FEL. In order to evaluate the performance of the undulator, magnetic field of the undulator was mapped by a detailed field measurement system. The field map obtained experimentally was compared with the map calculated by computer. The undulator field was found to be sinusoidal in the longitudinal direction and almost flat in the transverse direction. The performance of the undulator was found out to be sufficient to lase the JAERI FEL, because the irregularity of the field pattern of the undulator is sufficiently small.

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A long and slender position-sensitive 3He proportional counter with an outer diameter of 8mm and an effective length of 1.2m has been developed to measure neutron flux distributions in a critical assembly. Because it is difficult to make such a long and slender counter of con- ventional structure where an anode wire is stretched at the center of a cylindrical cathode tube, three prototype electrode systems of which anode wires are supported by insulator plates have been proposed. Among these three electrode systems, the electrode system of which anode wire is stuck on a ladder-shaped insulator has been adopted. The neutron flux distributions in the core of the Kyoto University Critical Assembly (KUCA) have been measured with the counter. Some non-uniformity of the detection efficiency has been observed throughout the counter. This non-uniformity has been corrected by calibrating the position-dependent detection efficiency with a gold wire activation method. A fairly satisfactory distribution, where the relative deviation from the gold wire activation method is 5.0%, has been obtained. Though the non-uniformity of the detection efficiency should be solved for precise measurements, the counter can be used for quick and rough measurements of neutron flux distributions.

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A plant information offering system was proposed to support operators in their selection and confirmation of the required information for plant operation under transient conditions in nuclear power plants. The system features include an automatic selection method for information and a dialog input method. The former selects plant information in response to plant status changes and operators' demands. The selection is performed based on the knowledge and data as struc- tured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. In the latter, both speech and CRT touch inputs are transformed into words in Japanese to realize an arbitrary input mode combination. The words are analyzed as a sentence before transforming them into a demand for related programs. A prototype system was evaluated using a BWR simulator, assuming abnormal transients such as loss of feedwater. The contents of the offered information were checked based on emergency operation guidelines. It was confirmed that appropriate information items are auto- matically selected in real time. Answers are generated in reply to the operators' demands. They include information added to reflect the plant conditions. As for dialog, simple and quick input is realized by combining speech and CRT touch according to the operating situation.

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This paper describes the experiments of the first Japanese nuclear ship'Mutsu', to investi- gate the effects of sea wave on the thermal hydraulics of marine reactor system while cruising through various sea conditions. The experimental data were analyzed in time-domain by RETRAN-02/GRAV code. This code was modified so as to simulate the ship motion effect on reactor thermal hydraulics. The data were also analyzed in frequency domain by Blackman- Turkey method for the calculation of the spectrum and response function. The experiments involving ship maneuvering were performed by cruising on different wave heights, as well as wave directions in the northern Pacific ocean. From the experiments, vertical acceleration due to ship motion was found to induce direct variation of water levels in the SGs and the pres- surizer. The water level variations were largest in the head wave, but smallest in the following wave. On the other hand, the following wave caused greater variation of the reactor power when the feed back control for the shaft revolution speed was used. Mechanism of response of water levels and reactor power with respect to the external forces are discussed. The response function (gain or phase shift) of reactor power to steam flow variation by the wave during cruising at rough sea condition was found to be roughly that without the work of control rod.

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Thermal striping is random temperature fluctuation generated by the mixing of flow streams at different temperatures. In a liquid-metal-cooled fast breeder reactor (LMFBR), it occurs in the upper plenum as a result of the mixing of sodium coolant jets from fuel assemblies and control rods. It is necessary to consider the temperature cycling in the context of fatigue on the component. A simplified model has been developed to analyze the attenuation of tempera- ture fluctuations from the flow streams to the structure surface. The analytical results were compared with experiments that simulated the upper core structure and impinging jets from fuel sub-assemblies and control rod channels in an LMFBR. Sodium and water were used as simulated fluids to examine the differences in their physical properties. The surface attenuation ratio decreased with the jet flow velocity. The ratio of surface temperature fluctuation to fluid temperature fluctuation is 30-80% in water tests and is 20-50% in sodium tests. These experimental attenuation values almost agreed with the values calculated by the boundary layer theory.

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Counter-current flow limitation (CCFL) is dominant phenomena for dryout in a debris bed which may be formed during a severe accident as observed in the Three-Mile Island unit-2. Actual CCFL situation in debris bed is very complex. It is difficult to treat the CCFL in the debris bed as it is. On the other hand, an annular flow model was developed to predict CCFL in a pipe by assuming a two-dimensional turbulent flow. If hypothetical flow channel were assumed for CCFL in the debris bed, CCFL in the debris bed could be treated with the same manner as for CCFL in a pipe. The purpose of this study is to investigate whether the annular flow model developed for CCFL in a pipe is applicable for CCFL in the debris bed or not. As the results, it is clarified that qualitative tendency of the CCFL in the debris bed consisting of larger particles than 3mm is estimated by the annular flow model developed for CCFL in a pipe, although the difference between the calculation and the data is large in higher and lower gas velocity. It is also clarified that wall friction factor calculated with the present analysis is twice to forth larger than that in the single phase flow through porous media.

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Single crystals of magnesium aluminate spinel (MgAl₂O₄) were heavily irradiated to fission neutron fluences from 1×10²⁶ to 2×10²⁷n/m² (E>0.1MeV) at 658 and 1, 023K in FFTF (Fast Flux Test Facility) to investigate their microstructural evolution under heavy neutron irradiation and to confirm their previously observed resistance to void swelling. At 658K no voids were observed up to 2.29×10²⁷n/m², but 1/4[110] interstitial loops were formed. These loops changed their habit planes from (111) to (110) with increasing neutron fluence. At 1, 023K the 1/4[110] type of interstitial loops grew to form stacking fault networks composed of stacking faults on each of the six equivalent {110} planes. Tiny voids were also observed to form preferentially on or near stacking faults after 1.37×10²⁷n/m². A limited number of very small cavities, con- tributing to a volumetric swelling of only 0.07%, were also observed in the crystal matrix after 2.17×10²⁷n/m². The possible mechanisms of suppression of void formation in MgAl₂O₄ are discussed.

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Fundamental investigations on preparation of Am(VI) in nitric acid solutions of various concentrations were carried out in order to develop a partitioning method for americium from high level liquid waste generated from the reprocessing of spent fuel. A mixture of ammonium peroxodisulfate and silver nitrate was used as oxidizing reagent. The yields of Am(VI) were determined as a function of the concentrations of nitric acid (0.01-2M) and oxidizing reagent (0.1-2M) and temperature (55-85°C). The yield of Am(VI) decreased with increasing concentration of nitric acid. In 1M nitric acid solution, increasing the concentration of ammonium peroxodisulfate resulted in increased of Am(VI) yield. When 1.5M ammonium peroxodisulfate and 0.01M silver nitrate were used, 90% of the americium in 1M nitric acid solution was oxidized to the hexavalent state at 85°C.

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Decontamination characteristics of urban building surfaces, based on the covering of clay suspensions, have been studied. The cation converting conditions of clay suspensions were determined by a flocculation series test. Contaminated specimens for test purposes were pre- pared by application of a radioactive solution which had been extracted from the soil of a 2km zone of the Chernobyl Nuclear Power Plant (ChNPP). The results of this study show that clay suspension technology was effective for decontamination of an urban environment. Especially, K-NH4 clay suspension shows good decontamination performance.

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A new type of passive integrating cup monitors was developed by using a 50mm radius stainless steel hemisphere to measured indoor radon (²²²Rn) and thoron (²²⁰Rn) concentrations. By placing a pair of cup monitors with an air exchange opening of diameter 5mm and four openings of diameter 20mm at a 20cm distance from wall during about three months, the con- centrations of both gases could be assessed from the α track densities on the cellulose nitrate (CN) films. The ²²²Rn and ²²⁰Rn concentrations were surveyed with the cup monitors in the different types of dwellings around Nagoya in Japan over three years. The ²²⁰Rn concentrations were rather high in the dwellings with soil wall, and the mean ²²⁰Rn concentration was 160±12 Bq•m^-3. The ²²⁰Rn exhalation rate from wall and the ²²⁰Rn diffusion were evaluated from the distribution of ²²⁰Rn concentrations in the dwellings. The results of the surveys have also clarified the relationship between the ²²⁰Rn concentrations at a 20cm distance from wall and the ²²⁰Rn progeny concentrations measured in the same dwellings. Then, the annual mean effective dose equivalent due to ²²⁰Rn progeny was expected to be 0.67mSv•yr^-1 in the dwellings with soil wall.

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This report provides a practical procedure to solve the diffusion synthetic acceleration (DSA) method in the three-dimensional discrete ordinate transport equation with the least computational time and hardware requirement. A study was made on the stability of the algorithm and it was found that the cause of the instability is strongly related to the coexist- ence of the plus and minus terms in the external source. A new procedure, plus and minus term separation, is proposed to ensure the unconditional stability, in which procedure the plus and minus terms are solved independently and then both solutions are added together to get the final solution. In order to enhance the convergence, an optimized SOR (successive over- relaxation) procedure was also investigated and applied to the method. This method provides a fully automatic optimization that ensures the best convergence in DSA process. The present method was encoded to a three-dimensional code TRITAC, and its efficiency was validated by numerical experiments on fast reactor calculations. The required number of Sn and DSA sweeps were reduced by 40% and 65% respectively by the present method.

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