In the present study, such mechanical properties of hydrogenated zirconium as sound velocities, elastic moduli, hardness and stress-strain curve have been evaluated by ultrasonic pulse-echo method, Vickers hardness test, tensile test and resonance method using cantilever beam specimens, and the influence of the hydrogen uptake on the mechanical properties of zirconium alloys has been studied. The elastic moduli of solid zirconium hydride estimated from the sound velocities were larger than that of pure zirconium metal, and slightly depended on the hydrogen content. The hardness of the solid zirconium hydride was much higher than that of the pure zirconium, and decreased with increasing the hydrogen content. The Young's modulus of the solid zirconium hydride obtained from the tensile tests was in good agreement with the value estimated from the sound velocities. The elastic moduli of hydrogen solid solution obtained by the resonance method decreased with the hydrogen content. The elastic moduli of partially precipitated hydride were also measured by the resonance method, and the results could be elucidated from the datum of the elastic moduli of the solid zirconium hydride and the hydride solid solution obtained in the present study.
Electronic structures of zirconium hydride and hydrogen solid solution have been evaluated by the X-ray photoelectron spectroscopy and first-principles molecular orbital calculation. From the valence band spectra of the solid zirconium hydride, the occurrence of the valence electron transfer from Zr 4d band to Zr-H bonding state was found to induce the reduction of Zr-Zr metallic bonds with increasing the hydrogen content, because of the formation of Zr-H covalence bonds. In the hydrogen solid solution, the 4d electrons were, as well as the hydrides, drawn off from the metallic bond onto the Zr-H bond, leaving less charge to participate in the Zr-Zr bonds. When hydrogen atoms reside in the zirconium metal, the electron transfer caused by the spatial distribution of H 1s electrons induces the charge depletion between the first-nearest zirconium atoms, which consequently weakens the surrounding metal-metal bonds. These results could elucidate qualitatively the datum of the elastic moduli of the solid zirconium hydride and the hydride solid solution obtained in our previous studies.
The fast reactor group constant set JFS-3-J3.2 based on the evaluated nuclear data library JENDL-3.2 has been widely used in fast reactor analysis. However, it was recently found that there were errors in the process of making the group constant and they were revised. This set is called JFS-3-J3.2R. In this report, effects of the errors on nuclear characteristics were evaluated by a comparison with a new reactor group constant set, JFS-3-J3.2R. This report shows that the errors mainly affect removal cross section and distort neutron spectrum. As a result nuclear characteristics, such as sample Doppler reactivity and reaction rate in a blanket region, are significantly affected. However, it is also shown that other characteristics, such as criticality and sodium void reactivity, are not affected because the effects of errors are canceled out as a total integrated result.
A design study of the power conversion system for the Gas Turbine High Temperature Reactor (GTHTR300) was carried out. The study aimed at reducing the total mass of main system components, which simplified system configuration by selecting the non-inter-cooled cycle, and improvement of the performance of power conversion components to enhance economics. The 3-dimensional aerodynamic design of the turbine and compressor achieved high polytropic efficiencies of 93 and 90%, respectively, while reducing the differential thrust of the turbo-compressor to 10kN as well as keeping a high surge margin of 30% for the compressor, which made it possible to attain a high power conversion efficiency of 45.8%. A horizontal turbo-machine layout, in which the turbo-compressor and generator rotors were connected by a diaphragm-coupling, was proposed to lessen the load requirements for magnetic bearings. The turbo-machine rotor, which passed over critical speeds of bending mode, fulfilled the standard limit of vibration amplitude of 75μm at the rated rotational speed by optimizing the stiffness of the magnetic bearings. The main focus of the heat exchanger design was size and mass minimization, while fulfilling the target temperature efficiency of 95%. The plate-fin type recuperator employed an off-set fin arrangement, having a square cross section of 1.2mm×1.2mm. The pre-cooler employed helicalcoil tubes with low lateral fins. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.
Japan Atomic Energy Research Institute (JAERI) has been conducting the design study of an original design concept of gas turbine high temperature reactor, the GTHTR300 (Gas Turbine High Temperature Reactor 300). The GTHTR300 is a greatly simplified HTGR-GT plant that leads to substantially reduced technical and cost requirements for earlier technology deployment. Also, it is expected to be an efficient and economically competitive reactor in 2010s due to newly proposed design features such as core design with two-year refueling interval, conventional steel material usage for a reactor pressure vessel, innovative plant flow scheme and horizontally installed gas turbine unit. This paper describes the original design features focusing on reactor core design, fuel design, in-core structure design and reactor pressure vessel design. In addition, a preliminary cost evaluation proved that the capital cost of the GTHTR300 is less than a target cost of 200 thousands Yen/kWe. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.
A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to capability of producing high temperature helium gas and its inherent safety characteristic. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved the rated thermal power of 30MW and reactor outlet coolant temperature 850°C in December 7, 2001. On March 6, 2002, "certificate of pre-operation test", that is, operation permit of the HTTR at rated operation mode, was conferred from the Ministry of Education, Culture, Sports, Science and Technology. For safe and steady execution of the test, the rise-to-power test was conducted step by step, that is, it was divided into three phases of the power levels of 10, 20, and 30MW. During the rise-to-power test, 22 tests were carried out such as calibration of nuclear instrumentation system to thermal power, performance of reactor control system, core physics, thermal expansion of high temperature components, radiation shielding performance, fuel and fission product behavior, performance at abnormal transient. All tests planned in the rise-to-power test had been successfully carried out and the performance of the HTTR was evaluated. This report describes the rise-to-power test results of the HTTR.
Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300) based on the experiences of development of the High Temperature Engineering Test Reactor (HTTR) of JAERI. Feasibility of the GTHTR300 fuel which is irradiated under high burnup condition of approximately 140GWd/t at the maximum was investigated. It was certified that failure of coated fuel particles during irradiation will not occur and the integrity of coated fuel particles will be kept. Cost evaluation for fuel production and fuel cycle was also studied in order to clarify the economical feasibility which is important item for the development of the GTHTR300. The fuel cost per electricity production is estimated to be less than 1.1Yen/kWh. The feasibility of the fuel cost is certified. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.
Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8×8, 9×9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPECTH-B Project). The high-burnup 8×8 fuel (average fuel assembly discharge burnup: about 39.5GWd/t), has been utilized from 1991. And the 9×9 fuel (average fuel assembly discharge burnup: about 45GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9×9 fuel assembly. Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9×9 fuel combined with the previously reported results of high-burnup 8×8 fuel. As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed.
The plant concept of 1, 700 MWe class next generation PWR (APWR+) has been established. This plant is developed as the improved version of APWR and is expected to be a main power generator since 2010s. The core consists of 257 17×17 fuel assemblies, which is the same as APWR. Although the fuel length was increased from 3.7m to 4.3m, the reactor vessel height was kept almost same as APWR by simplifying the core internal design. The safety systems are composed with active components as well as APWR, except that the systems were simplified by adopting heat removal through steam generators as a design basis. In this report, the utility requirements, which were established prior to the design evaluation, and the plant concept to realize these requirements are shown.
The distributed material database system named 'Data-Free-Way' has been developed by four organizations (the National Institute for Materials Science, the Japan Atomic Energy Research Institute, the Japan Nuclear Cycle Development Institute, and the Japan Science and Technology Corporation) under a cooperative agreement in order to share fresh and stimulating information as well as accumulated information for the development of advanced nuclear materials, for the design of structural components, etc. In order to create additional values of the system, knowledge base system, in which knowledge extracted from the material database is expressed, is planned to be developed for more effective utilization of Data-Free-Way. XML (eXtensible Markup Language) has been adopted as the description method of the retrieved results and the meaning of them. One knowledge note described with XML is stored as one knowledge which composes the knowledge base. Since this knowledge note is described with XML, the user can easily convert the display form of the table and the graph into the data format which the user usually uses. This paper describes the current status of Data-Free-Way, the description method of knowledge extracted from the material database with XML and the distributed material knowledge base system.
The probability of pyrochemical reprocessing in view of treatment and disposal of salt waste was studied. In considering the method of evaluating high level glass disposal, which is generated from wet reprocessing, we researched the certainty of vitrification of the two salt wastes, phosphate waste and surplus salt generated from the pyrochemical reprocessing. This report discovered that phosphate could be vitrified in fluoride phosphate glass and that the oxides converted from surplus salt by reacting with boric acid could be vitrified in borosilicate glass. It was found that both glasses are suitable for the present concept of radioactive glass disposal. Evaluating the engineering probability and the safety of glass after disposal, we confirmed the vitrification of the salt wastes from pyrochemical reprocessing as one of the promising methods for waste management and disposal.
In order to establish the extraction chromatography process for recovery of minor actinides from HLLW with a novel silicabased CMPO (octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide) adsorbent, some basic characteristics, such as dissolving behavior of CMPO from the adsorbent, thermal decomposition of the adsorbent and treatment method of organic wastes, were examined. It was found that the leakage of CMPO from the adsorbent in contact with an aqueous solution is the result of the solubility of CMPO in the solution. About 40-50ppm of CMPO constantly leaked into the effluent from the adsorbent packed column using 0.01M (M=mol/dm3) HNO3 as a mobile phase. The leakage of CMPO from the adsorbent could be effectively depressed with the utilization of the aqueous solution saturated by CMPO. TG-DTA thermal analysis results indicate that CMPO in the adsorbent decomposed at 200°C and the SDB-polymer at 290°C. The impregnated CMPO could be completely dissolved out from the support with acetone. Furthermore, the organic wastes such as CMPO, oxalic acid and DTPA those come from the elution procedure could be effectively decomposed with the Fenton reagent.
Non-destructive test of target plate model for fusion experimental reactor with defects between armor tile (25mm×25mm×10mmt) and cooling substrate (120mmL×25mm×25mm) were performed by Neutron Radio Graphy (NRG), X-ray Radio Graphy (XRG), infrared ray thermography and ultrasonic techniques. Following conclusions were derived; (1) From the result of heat transfer analysis, it is determined that breakaway defects over 2mm between armor tile and cooling substrate gives particularly serious effect on cooling function of the target plate. (2) Therefore, it is required to the non-destructive testing techniques with detection sensitivity in breakaway defect length between armor tile and cooling substrate less than 2mm. (3) Ultrasonic detection technique is not applicable to break away defect due to disturbance and damping of ultrasonic wave in armor tile and infrared ray thermography also gives no information of the defects. However, from the heat transfer analysis, it is suggested that thermography has a potential to detect over 3mm size defect by detecting temperature change difference due to defects. (4) Neutron radio graphy and X-ray radio graphy techniques exhibit very high sensitivity against less than 0.5mm defect and it is concluded that NRG method is most suitable non-destructive testing technique with high accuracy and reliability for the detection of whole body check of the target plate structure.
In this paper the concept of critical group is regulated following the historical flow and evolved to concrete application field. In the consideration of radiation protection to the member of public, the direct pickup of individual is scarce. In such a case, the concept of critical group is usually used, which is a concept of group handling of public radiation protection. Critical group is defined as a small group having homogeneous factors, i.e. age, dietary habits etc., which affect the estimation of radiation dose, and represent an individual who is expecting to receive the highest dose in the population. Critical group is usually used as an important concept in case of the estimations of nuclear facility siting and safety of radioactive waste disposal. This concept is used for dealing with a population from a part of the public radiation protection, coping with the concept of collective dose. As a result of investigation and discussion, the points and subjects are summarized as follows: (1) evolution of critical group concept in ICRP, IAEA etc., (2) concrete examples of critical group concept application, (3) especially, application of critical group concept in the field of radioactive waste management.
The technology of light water reactor will continue as a realistic energy technology for about half century at least from now on. The social agreement is necessary to continue nuclear power generation. Nuclear community must renew the conventional thought partially and have to approach the thought close to the value judgment in another social coordinate. "The Sociology of Science" on the atomic energy shall be such contents that be able to contribute to the unify or to attaches both coordinates as near as possible.