The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1, 000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society.
It is important to estimate internal exposure dose from intake radioactivity measured by human counter. The previous paper described application of the γ-γ coincidence method to human counter using two large HPGe detectors and showed equation to calculate activity without determination of γ-ray detection efficiency. This paper showed the results of further experiments using absorbers based on the previous fundamental experiments. When absorption of γ-rays in human body was simulated by depth of water on the source, the activities obtained for various depth agreed with true activity within ±10%. Furthermore, a moving source like a human in X-ray CT apparatus was measured to simulate whole body counting and the result showed usefulness of this method for human counter.
Technical review of conceptual multi power conversion system with productions of electrical power and hydrogen synthesis gas by methane partial oxidation reaction in combination of gas cooled reactor and SOFC was made. This system can produce electrical power, hydrogen synthesis gas and power load by coupling HTGR to supply high temperature thermal energy to cover enthalpy change in the methane reforming process. In present paper the performance of this system was discussed, and the following conclusions were derived. (1) This system requires very high performance electrolyte materials for SOFC in the high temperature service over 1, 000K, and CeSmO2 materials was doped by 3b elements such as Al3+, Ga3+ and Ga3+, and it is found that In3+ dopant improves density and ion conductivity of CeSmO2, and best composition of the doped material was optimized as Ce0.8Sm0.15In0.05O1.9. (2) In the methane partial oxidation test with SOFC single cell of Ce0.8Sm0.15In0.05O1.9, 74% free energy change was converted into electrical power and 90% methane reforming was achieved.
Numerical analysis has been performed for three-dimensional developing turbulent flow in the U-bend of strong curvature with rib-roughened walls by using an algebraic Reynolds stress model. In this calculation, the algebraic Reynolds stress model is adopted in order to predict preciously Reynolds stresses and boundary fitted-coordinate system is introduced as the method for coordinate transformation to set exactly boundary conditions along complicated shape in ribroughed walls. Calculated results of mean velocity and Reynolds stresses are compared with the experimental data in order to examine the validity of the presented numerical method and the algebraic Reynolds stress model. It has been pointed out as a characteristic feature from the experimental result that the maximum velocity appears near the inner wall of curved duct, which phenomenon is not recognized in mild curved duct. The present method could predict such velocity profiles correctly and reproduce the separated flow generated near the outlet cross section of curved duct. Adding to this, calculated results show clearly that the generation of maximum velocity near a inner wall is caused by pressure driven secondary flow which moves to inner wall from outer wall along symmetrical axis. As for the comparison of Reynolds stresses, the present turbulent model relatively predicts the experimental data well except for the flow separated region which is located near the outlet cross section of curved duct.
Rapid bursting tests using the Tube Rupture Simulation Test Rig (TRUST-2) and its overheating rupture analyses were carried out on the 2·1/4Cr-1Mo steel tube under an internal pressure load with extraordinarily high temperature condition. The tube failure behaviors were classified into three patterns of the ductile failure, the creep failure, and the ductile failure accompanied by creep, depending on the test conditions. In comparison between the test and the analysis on the creep failure and the ductile failure accompanied by creep, the time to failure of the analysis is estimated 25-50% shorter than that of the test, and the analysis is considered to be conservative. On the ductile failure, the times to failure of the test and the analysis become close but that of the analysis is still estimated about 10% shorter; the temperature to failure of the analysis is calculated about 100°C lower than that of the test; therefore, the conservatism of the analysis is confirmed. On the creep failure, if the time coefficient of the analysis is given about 1.5-2, the time to failure of the analysis becomes close to that of the test, which means that the analysis by use of the time coefficient 3 results in having the safety margin of 1.5-2 in the application to the Prototype FBR MONJU.
Nuclear Power Engineering Corporation (NUPEC) has been performing conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program, MISTRAL, that was devoted to measuring the core physics parameters of such advanced cores. The program consists of one reference UO2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. NUPEC has been analyzing the experimental results with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. This indicates that these applied analysis methods give the same accuracy for the UO2 core and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores.
Japan Atomic Energy Research Institute has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of a depressurization accident. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of th GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.
The purpose of this research is to clarify how psychological factors' impact on public acceptance of nuclear energy varies with people's degree of knowledge. For this purpose, we carried out questionnaire survey about nuclear energy at Suginami ward, Tokyo. After collecting data, we applied factor analysis to the data, and found 4 factors: Trust in the Authorities, Superiority of Nuclear Power Generation, Benefit for Nuclear Power Plants' Siting Areas, and Risk Perception about Nuclear Technology. In addition, using multiple regression analysis, we evaluated the impact of the 4 factors on 2 issues: the decision for or against nuclear policy and the reaction to nuclear power plant siting. Consequently, we found the change of impact of all 4 factors on the 2 issues. Especially, the impact of the 4 factors to the reaction to nuclear power plant siting was drastically changed by respondents' degree of knowledge.