The purpose of this research is to clarify how psychological factors' impact on public acceptance of nuclear energy varies with where they live and their degree of knowledge. For this purpose, we carried out questionnaire survey about nuclear energy at three urban areas and two nuclear power plant siting areas. After collecting data, we applied factor analysis to the data, and found four factors which construct cognitive structure of nuclear energy. Using multiple regression analysis, we evaluated the impact of the four factors on two issues: the decision for or against nuclear policy and the reaction to nuclear power plant siting, and compared changes of the impact by where respondents live and their degree of knowledge. Consequently, we found that the impact of all four factors on the two issues varies with where respondents live. We also found that the impact of respondents' degree of knowledge to four factors varies with where they live.
The purpose of this research is to clarify people's cognitive structure of nuclear energy, and to analyze how the cognitive structure varies with inhabiting areas, genders, and knowledge of nuclear energy. For this purpose, we carried out questionnaire survey of perception of nuclear energy in the urban areas and nuclear power plants (NPP) siting areas. After collecting data, we defined 8 categories in terms of respondents' inhabiting areas, genders, and knowledge, and applied factor analysis to each category's data. Consequently, we found 4 cognitive factors of nuclear: "trust in the authorities", "utility of nuclear power generation", "benefit for NPP siting areas", and "risk perception about nuclear technology", regardless of the respondents' inhabiting areas, genders, and knowledge. In addition, when the respondents assess many perceptions of nuclear energy, respondents living in urban areas tend to regard "trust in the authorities" as important, while respondents living in NPP siting areas tend to take into consideration of "risk perception about nuclear technology".
The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thoriumuranium fuel cycle. The notable feature of the MSR is that fuel salt flows the inside of the reactor accompanying nuclear fission reaction. In the previous study, the authors had developed numerical model to simulate the effects of the fuel salt flow on reactor characteristics and estimated the basic effects in a simplified system. This paper applies the model to the steady state analysis of the MSR system and simulates the effects of not only fuel flow but also fuel salt inflow temperature and residence time in the 1st loop system. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The following results are obtained: (1) the fuel salt flow affects the distributions of the delayed neutron precursors, especially long-lived one, and (2) the extension of residence time in the 1st loop system and the rise of fuel inflow temperature shows negative reactivity effects, decreasing neutron multiplication factor.
On-line and real-time estimation of time-varying reactivity in a nuclear reactor is necessary for early detection of reactivity anomaly and safe operation. Using a digital nonlinear H∞ estimator, an experiment of real-time dynamic reactivity estimation was carried out in the Very High Temperature Reactor Critical Assembly (VHTRC) of Japan Atomic Energy Research Institute. Some technical issues of the experiment are described, such as reactivity insertion, data sampling frequency, anti-aliasing filter, experimental circuit and digitalizing nonlinear H∞ reactivity estimator, and so on. Then, we discussed the experimental results obtained by the digital nonlinear H∞ estimator with sampled data of the nuclear instrumentation signal for the power responses under various reactivity insertions. Good performances of estimated reactivity were observed, with almost no delay to the true reactivity and sufficient accuracy between 0.05 cent and 0.1 cent. The experiment shows that real-time reactivity estimation for data sampling period of 10ms can be certainly realized. From the results of the experiment, it is concluded that the digital nonlinear H∞ reactivity estimator can be applied as on-line real-time reactivity meter for actual nuclear plants.
Temperature of heat transfer fluid flowing in latent heat storage medium could be kept near the melting temperature due to the latent heat. Therefore, latent heat storage technology has possibility to be used to reduce thermal load change of heat utilization system connected to HTGR (High Temperature Gas Cooled Reactor), which can improve reliability of the system including the reactor. This paper treated heat transfer characteristics of heat storage annular pipe as a simulation of heat exchanger pipe. Time dependent wall and gas temperatures and radial position of liquid solid boundary were studied by numerical and approximate analysis with thermal conductivity of phase change material (PCM) as a parameter. The analytical results were applied to estimate the possibility of the latent heat storage technique to reduce thermal load change for two heat utilization systems with 10MW connected to HTGR. When temperature difference of 100K from normal operation temperature occurred, a thermal load absorber with volume of about 15m3 can reduce 100K within 5K for three hours, which would be sufficient for operators to recognize the accident and to treat adequately. Therefore, latent heat storage technology has possibility for the application to thermal load absorber.
A new method was developed to assess the probabilistic effectiveness of severe accident management, including decision-making errors (DME), operation errors, and unavailability of accident management equipments (UOE). This method was applied to evaluated the risk decreasing effect at a typical 4-loop PWR plant with a dry containment vessel, and analyzed the sensitivity of DME and UOE. As a result, (1) after the preparation of accident management, the containment failure frequency was 1.0×10-8 (1/reactor year), and the risk decreased by 77%. In this case, it was shown that phase II accident management has 89% of risk decreasing effect to the containment failure modes concerned, and has sufficient effectiveness. (2) The results of the sensitivity study on DME revealed that the increase in risk due to a high-dependence cognition condition was about 4%, and the increase in risk due to loss of the double-check function in an organization was about 20%. (3) The results of the sensitivity study on UOE revealed that low pressure injections and recirculation systems were the most important in preventing containment failure. When the average time taken to repair the accident management equipment was tripled, the risk increased by about 13%. (4) This assessment method can be shown the positive/negative effect of accident management strategy clearly by using a logic operation formula, and is suitable when a third-party review is required.
In the present study, the phase diagrams for zirconium-oxygen-hydrogen ternary system were assessed between 573.15K and 1, 473.15K by means of CALPHAD technique. The calculated ternary isothermal section of zirconiumoxygen-hydrogen system at 973.15K was in good agreement with the experimental diagram. The calculated and experimental Sieverts' constants also accorded well. Isothermal sections of the other temperatures were also obtained, and drastic shifts of phase boundaries were prospected. From the zirconium (with hydrogen)-oxygen pseudo binary phase diagrams, it was predicted that dissolved hydrogen let the oxygen content for prior-β phase increase, and that the zirconium based fuel cladding became less ductile by the addition of hydrogen.
It is necessary to evaluate the influence of the colloid on radionuclide migration under geological disposal environment for safety assessment of radioactive waste disposal. In this study, batch sorption experiments and column transport experiments were carried out to evaluate the availability of analytical model of radionuclide migration based on assumption of equilibrated sorption among radionuclide, colloid and rock. Granodiorite was used in the experiments. Bentonite colloid sorbed by strontium was injected into the parallel plate made by the granodiorite. The comparison between the experimental and the calculated results of the column experiments showed the good agreements by using constant sorption ratio of tracer among the colloid, rock, and aqueous phase. This result indicates the availability of the model for the safety assessment. The flow rate of the column experiments was rather high compared with the natural groundwater flow rate, therefore the kinetic approach is still important to assess the colloidal transport and its effect on the radionuclide migration in the natural groundwater flow.
Three-dimensional numerical calculations have been performed on the generation of fluid driving force in liquidmetal magnetic fluid (liquid-metal including ferromagnetic particles) and water magnetic fluid (water including ferromagnetic particles) flowing in an increasing magnetic field. It has become clear from the calculations that the liquidmetal magnetic fluid including ferromagnetic particles with volume concentrations greater than 0.001 can produce the magnetic force which overcomes the Lorentz force and thus can generate the fluid driving force. Next, experiments have been performed using water magnetic fluid. These experimental results by the authors and others in literatures are compared with numerical results obtained by the present calculations. It has been proved that the present numerical calculations can predict nearly the pressure increase observed in the experiments if the magnetization of the magnetic fluid is given exactly.
Design study of a Gas Turbine High Temperature Reactor with electric power of approximately 300MW (GTHTR300) has progressed in Japan Atomic Energy Research Institute. The GTHTR300 is a simplified and economical power plant with a high level of safety characteristics and a high plant efficiency of approximately 46%. The nuclear, thermal and hydraulic design of the GTHTR300 is worked out under many kinds of limitations from economical and safety requirements. By applying an innovative refueling scheme using a sandwich shuffling method and longer effective burnable poisons, economical requirements of long refueling interval of two years in order to establish a high availability factor of over 90%, a high average fuel burn-up of 120GWd/t, and so on, were attained. Also a newly arranged control rod insertion scheme was adopted in order to meet the safety requirements, such as, a low control rod ejection worth of less than 0.2%Δk, the power density of less than 13W/cm3 and the maximum fuel temperature of lower than 1, 400°C. It was clarified by this study that the design to meet the economical and safety requirements is attainable. In this paper, the nuclear, thermal and hydraulic design features of GTHTR300 are described.
Japan Atomic Energy Research Institute (JAERI) has been developing the Gas Turbine High Temperature Reactor (GTHTR300) based on experience gained in development and operations of the High Temperature Engineering Test Reactor (HTTR) in JAERI. The basic fuel cycle concept in Japan is such that all spent fuel shall be reprocessed. Feasibility of the GTHTR300 spent fuel reprocessing was investigated so that the GTHTR300 can comply with the Japanese recycling policy. The Purex process was found to be essentially adaptable except for the head-end treatment. In the headend process, it was shown that carbon layers and graphite matrix around coated fuel particles are removed from a fuel compact by a burning method, and uranium can be taken out by destruction of the SiC layer with a hard disk crusher, followed by re-burning. Next, the Purex process can be supplied diluted by depleted uranium. To evaluate cost, a preliminary design of the head-end processing plant was studied and reprocessing unit price was evaluated. If the unit cost of waste disposal is assumed nearly equivalent to LWR's, the total fuel cycle cost of GTHTR300 was estimated to be about 1.58 Yen/kWh, which includes the reprocessing cost estimated at about 0.52Yen/kWh. The economical feasibility of GTHTR300 is thus confirmed. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.
A new methodology to construct distributed computing systems specially targeting nuclear power plant monitoring systems is proposed. In this framework, a monitoring system is composed of multiple modules and a client that administrates them. Each module is designed as a TTY-based program, and therefore has a great flexibility when it is developed. The client holds virtual modules, each of which works as an interface to a module in the remote hosts. Because the virtual modules are defined as a class in the meaning of object-oriented programming, the whole system is easily structured. A prototype of neural-network-based monitoring system has been developed utilizing this methodology, and the expected advantages have been confirmed.
This paper deals with liquid film dryout and rewetting behavior under BWR (Boiling Water Reactor) transient condition. The rewetting behavior has not been fully clarified. Therefore, the transient BT (Boiling Transition) tests were performed using a simulated BWR 4×4 rod bundle which many thermocouples were installed along one of the heater rod in order to clarify the re-wetting behavior. The following results were obtained; (1) The BT occurs initially just upstream of the spacer. The location of BT then propagates upstream, and finallyrewetting occurs upstream between adjacent spacers. (2) The liquid film front exists between spacers. (3) Rewetting propagation velocity decreases, as wall superheat increases.
Fuel cell seems to be one of resolutions for global warming so that large amount of feed hydrogen to fuel cell will be necessary. To meet the hydrogen demand, new hydrogen production system without CO2 emission will be required. The High Temperature Gas-cooled Reactor (HTGR) can produce high temperature helium gas that is around 950°C. Hydrogen production using HTGR is now being studied all over the world. The Japan Atomic Energy Research Reactor (JAERI) is planning the demonstration test of hydrogen production using the High Temperature Engineering test Reactor (HTTR). A design study on the steam reforming hydrogen production plant to be connected to the HTTR has been performed. In this report, interface technology between HTGR and hydrogen production plant is mainly discussed. In addition, safety requirements for hydrogen production plant are described. Flow scheme of the HTTR hydrogen production system is established to achieve stable controllability. Plant simulation analysis shows good performance of start up sequence. Component design of steam reformer and isolation valve for hot helium pipe that are special components in the HTTR hydrogen production system is presented. Heat and mass balance of hydrogen production plant is settled based on parametric survey of hydrogen production rate.
A program for tests on rotor dynamics was planned for the turbo-machine of the Gas Turbine High Temperature Reactor (GTHTR300). The rotor system of the turbo-machine consists of a turbo-compressor rotor and a generator rotor connected with a flexible coupling, each suspended with two radial magnetic bearings. The rotors, which are flexible rotors, pass over the critical speeds of bending mode. The magnetic bearing is required to have a high load capacity, about 10 times larger than any built thus far to support a flexible rotor. In the rotor design, the standard limit of the vibration amplitude of 75μm at the rated rotational speed of 3, 600rpm was fulfilled by optimizing the stiffness of the magnetic bearings. A test apparatus was designed to verify the design of the magnetic bearing suspended turbo-machine rotor of the GTHTR300. The test apparatus is composed of 1/3-scale test rotors, which are connected with a flexible coupling and driven by a variable speed motor. The test magnetic bearing was designed within the state-of-the-art technology to have a load capacity about 1/10 of that of the actual one. The test rotors were designed to closely simulate the critical speeds and vibration modes of the actual ones. This paper shows the test apparatus and the test plan for the magnetic bearing suspended rotor system. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.
Maintenance methods and procedures for the power conversion system of the Gas Turbine High Temperature Reactor 300 (GTHTR300) were examined. Because of being installed in the power conversion vessel which is primary pressure boundary and exposed to primary coolant contaminated with fission products, unique methods and procedures which are different from those for existing power plants and High Temperature Engineering Test Reactor (HTTR) are needed for especially the gas turbine and the compressor. The gas turbine-compressor assembly is pulled out from the power conversion vessel by a special trolley after generator vessel section is separated. Then, the gas turbine-compressor assembly is lifted and carried to a maintenance facility by a building overhead crane. During the removal process, preventive measures to reduce the exposure are taken by limiting working time and bagging the assembly to keep fission products from scattering. Before casings are open, the gas turbine-compressor assembly shall be in storage for a certain period to reduce the dose rate. In order to achieve a high plant availability, a spare gas turbine-compressor assembly is installed as an alternative to the assembly cooled in the storage. It was confirmed that, by applying those proposed methods and procedures, it is possible to keep the plant availability more than 90%. 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. Research and development of high temperature gas turbine plant and high temperature heat utilizing technology are now undergoing. The High Temperature Engineering Test Reactor (HTTR) is a research facility constructed by the Japan Atomic Energy Research Institute (JAERI). All rise-topower tests have been successfully carried out and the performance of the HTTR has been evaluated. Now, preparation for the operation with outlet coolant temperature of 950°C and safety demonstration tests are undergoing. This paper describes reprocessing technology of HTGR fuels. Coated fuel particles, consisted of a microsphere of low enriched UO2 with TRISO particles, are used as the HTGR fuels. In order to reprocess HTGR fuels, a head-end process is needed and JAERI had confirmed jet-grind method as basic technologies of the head-end process. Since Purex method can be used after the head-end process, a reprocessing system for the HTGR fuels could be established. Also the preliminary study on the methodology for disposing graphite blocks in a HTGR was carried out, and its evaluation results were briefly presented.
This paper is a lecture note on the continuous energy Monte Carlo method. The contents are as follows; history of the Monte Carlo study, continuous energy Monte Carlo codes, libraries, evaluation method for calculation results, integral emergent particle density equation, pseudorandom number, random walk, variance reduction techniques, MCNP weight window method, MCNP weight window generator, exponential transform, estimators, criticality problem and research subjects. This paper is a textbook for beginners on the Monte Carlo calculation.
Studies of accident and malfunction analyses are important because their conclusive results derived by engineering analyses are applied to nuclear power plant operation. In this paper, limiting the consideration to light water reactor, the factors that govern the design conditions of such reactors are first summarized. This is followed by an examination of the destruction mechanism of equipments and fittings such as piping, through analyses of such phenomena as cracking from highly-cycled thermal or mechanical fatigue, stress corrosion cracking, and fracture from corrosion and ductile straining, applying as basis the ASME technical standards and the thermohydrautic codes. Circumstances affecting instances of actual accidents and mulfunctions are then taken into account with the considerations required to be given from the engineering standpoint in applying the results of the present study to safety regulations and to practical plant operation.