Huge calculation time is required to evaluate neutron flux behind a shield in very large scale and complicated geometry such as a nuclear fusion reactor by Monte Carlo calculation. The density reduction method is proposed as a new methodology for the weight window generator on the variance reduction method to reduce calculation time in the present study. Calculation time can be drastically reduced by applying the density reduction method. It can be concluded that this method is very useful for the shielding calculation of very large scale and complex geometry such as a fusion reactor.

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  In the preparation process of a MOX powder for a MOX fuel fabrication facility, zinc stearate will be added into the MOX powder as an additional material to adjust the density. If too much of the material is added by mistake, the criticality characteristics of the MOX fuel will be influenced because of its neutron moderation effect. If a criticality condition should be induced by the excess zinc stearate, melting and pyrolysis of the material could be caused by the fission energy, and, therefore, the dynamic characteristics of the MOX fuel would be affected in a feedback mechanism. To provide a quantitative evaluation of such dynamics, thermal properties such as exothermic and endothermic calorific values, reaction rates, etc. were measured and correlated with the respective physical variations and release behavior of pyrolysis gas with the pyrolysis reaction. It was found that the exo/endothermic behavior of the material with rising temperature could be divided into six regions and that a rapid rise in pressure occurred over about 400°C caused by the pyrolysis reaction. Furthermore, on the basis of these results, an evaluation model for the thermal properties under criticality conditions was also investigated.

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  Reactor core status monitoring techniques have been developed to identify automatically the presence of a fuel assembly and double blade guide at each fuel exchanging step. This paper describes a method to identify the reactor core status using gray scale images obtained by a high resolution digital camera installed at the top of the reactor pool and test results obtained from images taken at an actual nuclear power generating plant. We developed a lattice mapping method for top guide positional data mapping on the reactor core images and a pattern matching method using the initial core image as the reference one. Evaluation testing of the system was carried out using more than 10 half reactor core images. We found the overall identification rate was about 96% for 1,825 image samples having the size of a fuel assembly. The processing time for identifying a half reactor core image including about 400 image samples was 60 to 90 seconds on a computer with 2.8 GHz CPU. This processing time is much smaller than the time for the shortest fuel exchanging interval.

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  Precise velocity measurement in a fuel rod bundle is required to improve the thermal hydraulic characteristics of Pressurerized Water Reactor (PWR) spacer grids. However, the use of conventional laser Doppler velocimetry (LDV) is restricted by the existence of invisible region in fuel rod bundles and insufficient spatial resolution for the narrow gaps in rod bundles. The new LDV was therefore developed to overcome these problems. The LDV is miniaturized with fiber optics embedded in a fuel cladding and can be inserted in an arbitrary grid cell instead of a fuel rod. The rod-embedded fiber LDV (rod LDV) is applied for cross-flow and axial-flow measurements at rod gaps and sub-channels to investigate turbulent velocity field in a rod bundle with mixing spacer grids. Distributions of mean velocity and turbulent intensity around the center rod of a 5×5 rod bundle are obtained. The measurements of this study are compared with particle image velocimetry (PIV) measurement for the same geometry and flow condition. Similarity of both measurements is confirmed qualitatively and some of data are in good agreement quantitatively.

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  Japan Atomic Energy Agency (JAEA) has been conducting an R&D work on the VHTR reactor system and IS hydrogen production system to realize hydrogen production using nuclear heat. As a part of this activity, JAEA is planning to connect an IS test system to the High Temperature Engineering Test Reactor (HTTR) to demonstrate its technical feasibility.
  This paper proposes a fundamental philosophy on the safety design of the HTTR-IS hydrogen production system including the methodology to select postulated abnormal events and its event sequences and to define safety functions of the IS system to ensure the reactor safety. Also the measure to clarify the IS system as non-reactor system is proposed.

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  Applicability of a leaching method to measurement of a clearance level of tritium in concrete was investigated. Static leaching tests were performed for tritium doped concrete samples in water, and the fraction of leached tritium was determined as a function of leaching time. The fraction increased rapidly in initial few days. After 10 days, the rate of increment was slow down, and stopped after 30 days in which the leaching ratio was 95±3%. In the case of actual concrete samples obtained from Japan Research Reactor No. 3, the leaching behavior was the same as that of simulated samples. The radioactivity of tritium determined by the leaching method showed good agreement with that from heating method. The method to leach out tritium from concretes with water is considered applicable to the analysis of a clearance level of tritium.

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  Previously reported “high sensitive detection of the fissile material in a solidification waste dram by 14 MeV neutron direct interrogation method” is the skillful method effecting the neutron moderation by the waste matrix itself. About this detection method, it is already reported to confirm and to be able to achieve an effective measurement for a concrete solidification waste. Moreover, even if it is only of the metal waste that the neutron moderation is not effective in the waste matrix, it was confirmed that positional sensitivity difference can be small and highly sensitive by using the additional moderator skillfully. In this report, it is examined that this detection method can be applicable effectively or not to a cloth and papers matrix waste filling within a dram. Here, in order to understand and to evaluate the detection characteristic by the difference of the filling density and the level of the position sensitivity difference, neutron transportation calculation of the model by assuming the amount of content cloth matrix waste (which depends on the filling density) as a calculation parameter was done. The method of reducing the positional sensitivity difference was also examined. As a result, it became clear that the detection limit of the natural uranium cloth matrix waste was reached to 0.1123 Bq/g and the positional sensitivity difference was reached in ±5%.

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  The Japan Atomic Energy Agency (JAEA) is constructing the Mizunami Underground Research Laboratory at Mizunami, Japan to establish general techniques for the assessment of the deep geological environment. The facility including two 1,000 m shafts and sub-stages at 100 m depths between two shafts is currently under construction in the sedimentary rocks. This study aims to evaluate the environmental changes around a large underground facility. To this end, hydrochemical changes in response to shaft excavation are assessed based on the observation of hydraulic head and groundwater chemistry around the facility.
  The observations indicated that rock formations with low hydraulic conductivity act as barriers to hydraulic disturbances, while higher conductivity zones provide a preferential flow path. Groundwater flow to the drifts creates chemical changes by mixing among chemically different groundwaters in higher conductivity zones. It is therefore meaningful to monitor the water pressure and chemistry at highly conductive rock formations during construction and operation of underground facilities. These investigations will provide the basic information on hydrochemical buffer capacity of the natural environment. Furthermore, observations suggest that grouting of conductive rock formations is important for maintaining the groundwater at near preconstruction levels so as to retain the buffer capacity of the rock formations used for safety assessment.

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  This technical note summarizes research activities on nuclear data carried out by Japanese Nuclear Data Committee (JNDC) during the fiscal years of 2003 and 2004. During this period, the nuclear data files for special purposes (JENDL-HE-2004 and JENDL-PD-2004) were released. Other activities are described: analysis of post nuclear fuel irradiation experiments, nuclear chart and nuclear data evaluation for astrophysics.

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