Monte Carlo machine, Monte-4 has been developed to realize high performance computing of Monte Carlo codes for particle transport. The calculation for particle tracking in a complex geometry requires (1) classification of particles by the region types using multi-way conditional branches, and (2) determination whether intersections of particle paths with surfaces of the regions are on the boundaries of the regions or not. using nests of conditional branches. However, these procedures require scalar operations or unusual vector operations. Thus the speedup ratios have been low, i.e. nearly two times, in vector processing of Monte Carlo codes for particle transport on conventional vector processors. The Monte Carlo machine Monte-4 has been equipped with the special hardware called Monte Carlo pipelines to process these procedures with high performance. Additionally Monte-4 has been equipped with enhanced load/store pipelines to realize fast transfer of indirectly addressed data for the purpose of resolving imbalances between the performance of data transfers and arithmetic operations in vector processing of Monte Carlo codes on conventional vector processors. Finally, Monte-4 has a parallel processing capability with four processors to multiply the performance of vector processing. We have evaluated the effective performance of Monte-4 using production-level Monte Carlo codes such as vectorized KENO-IV and MCNP. In the performance evaluation, nearly ten times speedup ratios have been obtained, compared with scalar processing of the original codes.

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Minor actinide (MA) transmutation performances were evaluated for BWR cores with mixed oxide (MOX) fuels of various hydrogen-to-heavy metal atom number ratios (H/HM) to decrease the work for high-level radioactive wastes (HLW) management. One effective approach to increase the MA transmutation ratio is a multi-recycle core of MOX fuel with MA. The decrease of the MA-to-fissile Plutonium (Puf) amount ratio makes it possible to recycle MA without additional plutonium (Pu) obtained from reprocessing of LWR fuel. For this purpose, a T-ratio, which is the MA-to-Puf amount ratio change during burnup, of less than unity is required, and H/HM must be less than 3 when MA enrichment is more than 2wt%. For multi-recycle cores with the H/HM of 3 and the initial MA enrichment of 5wt%, more than 50% of MA load initially can be transmuted by the third recycle.

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In case of a shielding analysis of the geometry having thick and complicated structures with a Monte Carlo code, it is a serious problem that it takes too much computer time to obtain results with good statistics. Therefore, it is very important to reduce variances in the calculation. In this study, a method to determine the importance function in 3-dimensional Monte Carlo calculation with geometry splitting with Russian roulette was developed for the shielding analysis of thick and complicated core shielding structures. Only two essential importance ratio curves for one material enable us to determine the importance function easily in the shielding calculation.
The validity of this method was confirmed through a simple benchmark calculation. From the comparison with the result obtained by using weight window (W-W), it was shown that the present method can give an accurate result on the same level with W-W method with less trial and errors. And this method was applied to an actual reactor core shielding analysis to confirm its applicability to a 3-dimensional thick and complicated structure.
Using this method, the variance reduced calculation can be easily realized with the developed importance determination procedure, especially in case that parameter survey calculations are required in order to determine the shield thickness in a design work of a thick and complicated structure. Accordingly, it became easier to use Monte Carlo method as a powerful tool for a reactor core shielding design.

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This paper describes the results of a transient experiment using low enriched (<20wt% ²³⁵U) uranium silicide research reactor miniplate fuel. Three silicide miniplate fuel specimens assembled in a triplet configuration with a water gap of 2.38mm were pulsed in the Nuclear Safety Research Reactor (NSRR) with an energy deposition of 78cal/g•fuel.
The results obtained in this study are summarized as follows:
(1) Under a restricted condition of coolant cross flow, the mid-plate in the triplet configuration (pulsed at 78cal/g•fuel) showed an extremely high peak cladding surface temperature (PCST), 475°C, which was higher than that of the other two plates (173 and 192°C). The temperature of coolant in the water gap increased rapidly to 140°C while that at the non-gap locations increased to about 50°C. Despite the poor cooling conditions, especially at the mid-plate, the three silicide miniplate fuels did not fail mechanically. Owing to the comparatively slow quenching rate from PCST and/or the small magnitude of the temperature drop during quenching, the fuels might not be subjected to the local tensile stress necessary to cause through-plate cracking.
(2) The bowing of the miniplate fuel caused the 2.38mm water channel gap to be narrowed to 48% of its maximum. The uneven coolant temperature profiles in the water gaps between the miniplates might be a cause for the bowing.

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An analytical assessment is made of the potential effects of irradiation-induced transient creep on the behavior of the TRISO-coated fuel particles of the New Production Modular High Temperature Gas-Cooled Reactor (NP-MHTGR). An analytical solution is presented for the three-layer particles, which includes transient creep in addition to steady-state creep behavior. The solution allows for evaluating the effects that transient creep has on individual particle stresses and for determining failure probabilities for particle batches using the Monte Carlo approach. Because experimental data needed to determine parameters for a transient component in a creep model for the pyrocarbons is not available, a range of possible parameter values were considered in the assessments. It was shown that transient creep measurably affects particle stresses early in the irradiation life of the particle. At that time, the hoop stress in the primary load bearing layer of the particle is in compression and the particle is not vulnerable to pressure vessel failure. Later in irradiation, the effects of transient creep were typically shown to be less significant. Thus, transient creep had less than an order of magnitude effect on batch failure probabilities for prototypical NP-MHTGR fuel particles and was much less significant than steady-state creep. Whether the presence of transient creep increased or decreased the particle failure probability was dependent on the specific values used for the transient creep material properties.

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In order to develop a feasible process for production of enriched ¹⁵N, nitrogen isotope separation by ion exchange has been studied. The attention has been placed on the concentration of the eluent LiOH solution introduced into the ion exchange column, packed with specially synthesized cation exchange resin. The ammonium ion adsorption band initially charged in the resin bed was eluted in a reverse breakthrough manner. The separation coefficients ε have been confirmed not to be seriously affected by the eluent concentration. However a tendency of slight decrease in ε has appeared in the high concentration region. The values of HETPs of the process have been also shown almost constant, in spite of the large difference in the migration velocity depending on the eluent concentration. The results suggest that the ion exchange rates are very fast at high concentrations of the eluent, probably due to the large mobility of neutral ammonia in the resin.

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Bauxite wastes of alumina manufacture, i.e., red muds, have been tested for radiocesium and strontium removal from water. The red muds were water-washed, acid-, and heat-treated before usage to produce hydrous oxide like sorbents. Surface treatment of the sorbent was benefical for ¹³⁷Cs uptake, while heat-treatment was detrimental to the -SOH surface sites responsible for high ⁹⁰Sr affinity. Fractionation of the sorbent with respect to apparent grain size did not produce significant differences in the sorption efficiency. The distribution coefficients vs. equilibrium activity in solution showed a maximum with Cs, and a gradual decrease trend with Sr. The solution activity vs. adsorption data were fitted to B. E. T. (essentially types IV-V) isotherms for Cs and B. E. T.-Langmuir isotherms for Sr. Desorption, temperature-, pH-, and ionic strength-dependence tests revealed that the primary mode of sorption for both cations is specific adsorption while the secondary mode is ion exchange. A rise in pH favours the ion-exchange sorption of Sr while the specific adsorption of Cs is negatively affected. Competitive adsorption of an inert electrolyte, i.e., NaCl, severely hinders Cs sorption, while Sr sorption on water-washed red mud is not significantly affected.

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The behavior of photochemical reactions of Np(V) in a nitric acid solution was studied under various test conditions as follows. As light sources, a mercury lamp and a semiconductor laser were used, and the light from the sources was irradiated on 2ml of neptunium nitric acid solutions in a quartz cell. The variables in these photochemical tests were the light irradiation rates (0.2 and 1.5W/cm²), the wavelength regions of the irradiation light (ultraviolet (UV) (250-400nm)), visible (400-600nm) and 980nm), the acidities of the nitric acid (1 and 3_M) and the kinds of addition reagents (hydroxylamine+hydrazine and urea). The photochemical changes in Np valences in the test solutions depending on the irradiation times were analyzed using a spectrophotometer. Based on these tests, the following results were obtained.
• Neptunium(V) was photooxidized to Np(VI) by the irradiation of UV light on a nitric acid solution.
• The higher the irradiation rate and the concentration of nitric acid, the easier the photo-oxidation reaction of Np(V) progressed.
• The addition reagents influenced the oxidation reaction rate and the degrees of the progress in the photooxidation reaction of Np(V)→Np(VI).
Based on the results of these tests, the quantitative control of the photooxidation reaction of Np(V) was found to have much potential for a valence adjustment by selecting the appropriate conditions of the UV irradiation rates, the acidity of HNO3 and the kinds of reductants.

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A method is proposed to assess the effect of nuclear education. In this method, the nuclear education is treated as a part of the activities for public acceptance (PA), and a unit PA activity is assumed to give the same effect on the public, in essence, as a unit of nuclear information given by the newsmedia. Moreover, the change of attitude to nuclear energy is assumed to originate from enhanced understanding which, in turn, is brought by the stimulus given by the nuclear education. With the values of constants determined by using the data in Japan, example calculations were made for the educational time b₀ and the infiltration rate of education into minors B as parameters. It became clear from this calculation that the attitude to nuclear energy formed in the age of school children plays an essential role in shaping future public opinion since it is held in individuals without any notable modification for a long time after its formation, and that the effect of nuclear education to minors emerges depending on the variables b₀ and B in a highly non-linear manner. It was also found that there exists an optimum condition for nuclear education to attain the maximum amelioration of public opinion under a given condition of man-power for educational workers.

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In a course of a design study of the JAERI passive safety pressurized water reactor (JPSR), a complete loss-of-flow transient caused by a trip of all pumps was analyzed with the RETRAN code to determine an inertia of canned-motor pump utilized as the primary coolant pump and to confirm feasibility of the design condition. This transient was selected because the pump had a low inertia rotor inducing fast flow coastdown, and among the transients in which the pump had dominant effect on the departure from nucleate boiling (DNB), the analyzed transient was severest in view of the DNB occurrence. The DNB threshold was related, based on sensitivity calculations, with the coolant density reactivity coefficient and the pump inertia. From the calculations, it was concluded that the pump inertia higher than 250 kg•m² (8% of the ordinary PWRs) was necessary for preventing the DNB occurrence for the present design of JPSR, regardless of the actuation of the reactor scram. The DNB occurrence could be prevented only by the inherent nature of the reactor core which reduced the power by insertion of negative coolant density reactivity during the transient and this was one of major features of JPSR. It was shown by a rough estimation that the necessary condition could be practically realized by incorporation of a cylindrical-type flywheel.

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The WAVE experiments have been performed at JAERI to investigate the CsI deposition onto the inner surface of pipe wall under typical severe accident conditions. It was shown that relatively large amount of CsI was deposited at the upstream floor of the pipe and that larger amount of CsI was deposited on the ceiling than the floor at the downstream. Analyses of the experiments have also been conducted with the three-dimensional thermohydraulic code, SPRAC, and the radionuclide transport analysis code, ART. The experimental results were well reproduced with ART by using peripherally subdivided pipe cross section and associated representative thermohydraulic information from SPRAC prediction. It was clarified through the present experiment and analyses that major deposition mechanisms for the chemical form of CsI are thermophoresis and condensation. Accordingly, the coupling of the FP behavior and the detailed thermohydraulic analyses was found to be essential in order to accurately predict the CsI deposition in the pipe, to which little attention has been paid in the previous studies.

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The results of the subcriticality measurements by various techniques performed for the UTR-KINKI reactor, a light-water-moderated and graphite-reflected coupled-core reactor, are presented. The result of the source-multiplication measurement indicates that the apparent dependence of the subcriticality on the detector position is significantly observed even under a condition 2$ subcritical, and that the location of neutron source remarkably influences the subcriticality obtained. On the other hand, the spatial dependence in the source-jerk measurement is slight relative to that in the source-multiplication and the rod drop measurements. Furthermore, the result of the Feynman-α measurement suggests that much more samples should be acquired for the reduction of the experimental uncertainty to the same level as that in the above measurements.

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