This paper analyzes the view expressed by the Japanese government on the role of nuclear energy for energy security through scrutiny of Japan's policy documents. The analysis revealed that the contribution by nuclear energy to Japan's energy security has been defined in two ways.
Nuclear energy improves short-term energy security with its characteristics such as political stability in exporting countries of uranium, easiness of stockpiling of nuclear fuels, stability in power generation cost, and reproduction of plutonium and other fissile material for use by reprocessing of spent fuel.
Nuclear energy also contributes to medium- and long-term energy security through its characteristics that fissile material can be reproduced (multiplied in the case of breeder reactor) from spent fuels. Further contribution can be expected by nuclear fusion.
Japan's energy security can be strengthened not only by expanding the share of nuclear energy in total energy supply, but also by improving nuclear energy's characteristics which are related to energy security. Policy measures to be considered for such improvement will include (a) policy dialogue with exporting countries of uranium, (b) government assistance to development of uranium mines, (c) nuclear fuel stockpiling, (d) reprocessing and recycling of spent fuels, (e) development of fast breeder reactor, and (f) research of nuclear fusion.

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An experimental study of developing flow in a rectangular duct with arrays of blocks was carried out at the Reynolds number of 3.26×10⁴ by using a Laser-Doppler velocimeter. The total amount of ten blocks are staggered each other and upstream tangent with rectangular cross section are set to obtain fully developed turbulent flow. The test section duct with blocks was placed at distance of 40.5 times hydraulic diameter from the duct inlet. Three components of mean velocity and the five components of the six Reynolds stresses were measured at nine different locations to clarify the development of turbulent structure. The measurement has been performed in the flow region between number 7 and number 8 blocks. As a result of this experimental study, it was pointed out as a characteristic feature that the two peaks of streamwise velocity were generated in the central and wall side regions respectively. The measurements of secondary flow vectors suggest that this phenomenon is caused by transforming lower velocity fluid to upward of block from leading edge of block and the large values of secondary flow, which reach maximum percentage of 50% for mean bulk velocity, are produced at leading edge and wake region of block. Adding to this, separated and reattached flows are observed in wake region of blocks and the distributions of three normal stresses indicate a strong intensity near the reattachment region. The distributions of Reynolds shear stress show the opposite sign region, which is owing to streamwise velocity distorted by secondary flow.

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High resolution TEM observations on surface oxide layers and uncracked grain boundaries intersected by the surface were carried out on alloy 600 exposed to simulated primary water environment of pressurized water reactors. A focused-ion beam micro-processing has been applied to prepare electron transparent foils for the cross-sectional surface observations. The surface oxide layer contained two different layers. The outer layer consisted of granular NiFe₂O₄ covered by NiO. The inner layer consisted of Cr rich oxides such as Cr₂O₃ and (Cr, Fe)₃O₄ and a metallic Ni phase. It is suggested that selective oxidation of chromium occurs in the inner layer. The observations that oxygen penetrated along the uncracked grain boundaries intersected by the surface indicated a larger diffusivity of oxygen in the stressed gain boundaries than that in the matrix. These results suggested the following model for the growth of the oxide double layer: the outer layer is formed by a metal dissolution and oxide precipitation mechanism, and the inner layer is formed by an oxygen ion diffusion mechanism.

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A model has been developed for prediction of pore water chemistries and solid constituents of cement, in order to study the cement degradation process caused by dissolution of hydrates. The dissolution models of portlandite, hydrated calcium-silicate gels and other hydrates, which are the main hydrates in cement, are introduced into the geochemical code to couple with the one-dimensional advective-dispersion transport model. The pH and the chemical concentrations in pore water, and the distributions of pore volume and elements in solid are studied by using this coupled code. The calculated results agree with the permeability test results for hardened cement. By using calculated results of distributions of elements, hydrates and pore volume in solid, the dissolution process in which the hydrated cement is altered by its own dissolution is discussed.

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Conventional nuclear reactors converge energy produced by fission to thermal energy of point nuclei, and subsequently to electric power by means of turbines. The conversion efficiency of fission energy to electric power in these systems has not yet gone beyond thirty-odd percent, and almost all applications are restricted to power generation. On the other hand, the collision theory since Rutherford's study indicates that the energy of a charged particle is transferred preferably to an electron with light mass by collision, except for the case of a low velocity charged particle where energy is transferred to a point nucleus with heavy mass. By using electrodynamics this paper shows the possibility that energy of a fission fragment can be transferred preferably to an electron in an atom such as rare gas by collision, and that the emitted energy of a secondary electron is transferred preferably to an electron in a rare gas atom by excitation caused by the collision. Thus, as a result of the photon emission from excited electrons in the rare gas, nuclear energy can be converted to photon energy. The photon energy emitted and the transition probability are derived by the first-principle electronic state analysis. Applications of this technology may possibly lead to the construction of new and various photoindustries.

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The fuel addition method or the neutron absorption substitution method have been used for determination of large excess multiplication factor of large sized reactors. It has been pointed out, however, that all the experimental methods are possibly not free from the substantially large systematic error up to 20%, when the value of the excess multiplication factor exceeds about 15%Δk. Then, a basic idea of a revised procedure was proposed to cope with the problem, which converts the increase of multiplication factor in an actual core to that in a virtual core by calculation, because its value is in principle defined not for the former but the latter core. This paper proves that the revised procedure is able to be applicable for large sized research and test reactors through the theoretical analyses on the measurements undertaken at the JMTRC and JMTR cores. The values of excess multiplication factor are accurately determined utilizing the whole core calculation by the Monte Carlo code MCNP4A.

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In order to remove carbon dust and fission products (FP) submicron particles released from incore of 600 MWt High Temperature Gas cooled Reactor system (GTHTR300; Outlet temperature/pressure, 1, 173K/6MPa), a new high temperature FP filter media (thickness=1.6 mmt) was made of two grain size of Hastelloy-X (13 and 20μm) on the operation conditions (High temperature strength at 1, 273K>1.5MPa, removal efficiency≥90%, initial pressure drop≤0.4%) and tested its filtration performance.
With these results, fundamental specifications of FP filter for HTGR-GT were estimated and the following conclusions were derived.
(1) New filter materials exhibits very high tensile strength over 3MPa at 1, 273K.
(2) 90% removal efficiency against submicron Ag particle can be achieved by coarse grained filter with 20μm.
(3) 0.4% low pressure drop in HTGR-GT system can be achieved by coarse grained filter with 20μm.
(4) On the base of the results obtained in present study, fundamental specifications of FP filter were estimated as following;
Particle size of Hastelloy-X powder: 20μm, Filling factor: 60%, Surface flow rate: 6m/min, Initial pressure drop: 0.4%, Removal efficiency: 90%, Filter thickness: 1.6mm

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The R & D of Seismic Emergency Information Transmission System has been conducted involving the latest progress in earthquake engineering with regard to estimation techniques on the hypocenter, fault and earthquake motion parameters and in information technologies. This system is the disaster management system which consists of user site and disaster information center and is capable of mutual information transmission through Inter-Net and walkie-talkie. The concept of the disaster management system which is adaptable with DiMSIS (Disaster Management Spatial Information System) developed by professor Kameda et al. of Kyoto University has been established. Based on this concept, a prototype system has been developed. This system has following functions, (1) compatible application both in usual condition and emergency time, (2) the decentralized independence, and (3) the integration of space and time information. The system can estimate the earthquake motion information with 500m square mesh in a local area and transmit in a few minutes. In the development of the system, seismometer network, surface soil database and amplification functions were prepared for the examination of system function. Demonstration against the Tokai area was carried out and the function was verified.

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If a severe accident occurs in a pressurized water reactor plant, it is required to estimate dose values of operators engaged in emergency activities such as accident management, repair of failed parts. However, it might be difficult to measure radiation dose rate during the progress of an accident, because radiation monitors are not always installed in areas where the emergency activities are required.
In this study, we analyzed the transport of radioactive materials in case of a severe accident, investigated a method to obtain radiation dose rate in the plant from estimated radioactive sources, made up a prototype analyzing system from this design study, and then evaluated its availability. As a result, we obtained the following: (1) A new dose evaluation method was established to predict the radiation dose rate at any point in the plant during a severe accident scenario. (2) This evaluation of total dose including access route and time for emergency activities is useful for estimating radiation dose limit for these employee actions. (3) The radiation dose rate map is effective for identifying high radiation areas and for choosing a route with lower radiation dose rate.

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The purpose of this study is to develop oxygen sensor which can measure the oxygen potential of the fuel in a nuclear reactor. The oxygen sensor with CaO stabilized zirconia solid electrolyte has been specially designed in order to prolong its life time. Electromotive force (EMF) of solid galvanic cell Ni/NiO |ZrO₂-CaO|Fe/FeO was measured in both the outpile tests and the in-situ tests using Japan Material Testing Reactor (JMTR), and the characteristics of EMF was discussed.
In the out-pile test, it was found that the EMF was almost equal to the theoretical values at temperatures ranging from 700 to 1, 000°C, and the life span of the sensor was very long up to 980h at 800°C. In the in-situ test, it was found that the EMF showed almost the reliable values up to 300h (neutron fluence (E>1 MeV) 1.5×10²³n/mm²), at temperatures from 700 to 900°C. The imprecision of the EMF was found to be within 6% of the theoretical values up to 1, 650h irradiation time (neutron fluence (E>1MeV) 8.O×10²³n/m^-2) at 800°C, The oxygen sensors were found to be applicable for the oxygen potential measurement of the fuels in a reactor.

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With a view to utilizing in post-irradiation examination (PIE) of fuel assemblies the powerful technique of nondestructive X-ray computed tomography (X-ray CT scanning) widely used in medical practice, the technique was further developed and successfully applied to a spent fuel assembly irradiated to 59 GWd/t in the JOYO experimental FBR.
The effect of gamma rays emitted from the spent fuel was mitigated by the use of pulsed high-energy X-ray source, which contributed to obtaining clear cross sectional images, from which a 3-dimensional image could be derived.
Application of the technique thus developed should permit enhancing the resolution of measurements and simplifying fuel PIE. Defect size was measured within an error of ±0.3mm, and differences in density distinguished down to ±4% in a material of density -8g/cm³ density.
With such accuracy, non-destructive PIE could well come to replace destructive method in many instances, to enhance efficiency and reduce radioactive waste. The technique could further be applied to the analysis of overall fuel assembly behavior, such as bundle-duct interaction and central void distribution affected by irradiation temperature.

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Ordinary monopolar electrolytic decontamination needs manual handling to connect an anode of the current power supply directly to the radioactive metallic wastes. The procedure results in radiation exposure for operators. The purpose of this study is to develop a bipolar electrolytic decontamination method in which no connection of the current power supply to the radioactive metallic wastes is needed.
Engineering scale decontamination equipment was assembled to validate the method using up to 500mm long samples. Test results showed that the piping was dissolved more effectively than plane samples because of lower leak current. Moreover, dissolution fluctuation is almost homogeneous (in the standard deviation of less than 10%) for plate and pipe sample. The fluctuation was larger at the bending plate sample. It was also evaluated that the samples could be decontaminated remotely even in a basket and that safety assurance and radiation exposure reductions due to the remote handling were feasible.

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Public perception on safety is the key factor for achieving public acceptance of the high-level radioactive waste (HLW) disposal program. Past studies on public perception and HLW management have confirmed that the pubic do not share the confidence of the experts in safety and feasibility of HLW disposal. The importance of a more comprehensive approach to enhance acceptability of the HLW disposal technology is recognized.
This paper proposes a framework for inducing the implementers and regulators to improve compatibility of the HLW disposal technology with social values. In this framework, the implementers and regulators identify technical components which are subject to substantial influence from public concerns. Then, they manage these components through the following actions: (1) establishing policies, targets and plans to make these components compatible with social values, (2) developing and utilizing the components based on the above policies, targets and plans, (3) cheking the extent of compatibility through intensive risk communication and (4) improving the process of developing and utilizing the components. This framework requires information disclosure and evaluation by an independent body which are expected to intensify the incentive to take the above actions. Canada's environmental assessment review process regarding the HLW disposal concept suggests that this framework could work effectively.

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[in Japanese]

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[in Japanese]

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