The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395°C and coolant outlet temperature of 850°C/950°C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. The reactivity insertion test demonstrates that rapid increase of reactor power during withdrawal of the control rod is restrained by only the negative reactivity feedback effect without operation of the reactor power control system, and the kemperature transient of the reactor is slow. A one-point core dynamics approximation with one fuel channel, one fuel temperature coefficient and one moderator coefficient could not simulate accurately the reactor power behavior. On the other hand, an original new method using some temperature coefficients for regions in the core was very effective. It is crucial to evaluate this method precisely to simulate a performance of HTGR with showing excellent inherent safety features under the reactivity initiated incident by control rod withdrawal error. Moreover, all experimental data of reactivity insertion tests and analytical results are released.
Intermittent and heterogeneous behavior caused by interface motion is important for gas-liquid two-phase flow dynamics. Its complex characteristics are commonly treated with constitutive equations obtained from experimental database in the conventional two-fluid model. Numerical simulation by the extended two-fluid model, developed to calculate interface motion directly, can be an alternative tool to a part of experiments. Correlation of CCFL, counter current flow limiting, of upper tie plate is important for coolability assessment of BWR fuel assembly during LOCA conditions. Airwater CCFL experiments of a perforated thick plate were performed, and the falling water flow rates were measured as a function of upward air supply. Three-dimensional simulation of the experiments using the extended two-fluid model predicted the observed CCFL characteristics within data scatter. It is expected the simulations by the extended two-fluid model is useful to actual tie plate design of future reactor fuels.
Fluorination of uranium and plutonium dioxides by fluorine is discussed for the flame tower reactor, which is one of possible reactors in FLUOREX process, a hybrid reprocessing combining a fluoride volatility process and a solvent extraction. A reaction model for a shrinking particle with an un-reacted shrinking core is applied to the fluorination of plutonium dioxide and the reaction rate constants are determined in low temperature range around 400°C using the reported kinetic data [Iwasaki et al., J. Nucl. Sci. Technol., 11, 403 (1974)]. The determined rate equations are used to analyze the fluorination of uranium and plutonium dioxides in the high temperature range (800-1, 200°C), where the flame tower reactor is operated. For uranium dioxides the diffusion of fluorine to the surface of the particle controls the overall fluorination reaction, and the model can predict the recovery of uranium to the product of uranium hexafluoride. For plutonium dioxides, however, the recoveries of plutonium calculated by the presented model do not agree with the experimental recoveries. It is suggested that the reaction mechanism for the fluorination of plutonium dioxide in the high temperature range is different from that in the low temperature range.
Japan Atomic Energy Research Institute (JAERI) has been developing a graphite moderate and helium cooled High Temperature Gas-cooled Reactor (HTGR) with gas turbine, the GTHTR300 based on experience gained in development and operations of the High Temperature Engineering Test Reactor (HTTR) in JAERI. The GTHTR300 is a simplified and economical power plant with a high level of safety characteristics and a high plant efficiency of approximately 46%. Cost evaluation for plant construction and power generation is studied in order to clarify the economical feasibility of the GTHTR300. The construction cost is estimated to be about 200 thousands Yen/kWe. The power generation cost is estimated to be about 3.8Yen/kWh by the conditions of 90% load factor and 3% discount rate. The economical feasibility of the GTHTR300 is certified. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.
To establish the reliability evaluation method for nuclear components, we developed a probabilistic seismic hazard evaluation code SHEAT-FM (Seismic Hazard Evaluation far Assessing the Threat to a facility site-Fault Model) using a seismic motion prediction method based on fault model. The seismic motion prediction method is usually used for defined fault. In this code, we propose to apply this method for undefined fault in the seismic hazard evaluation. This report describes the outline of SHEAT-FM code and sample problem for a model site, and result of the comparison of seismic hazard curves using fault model and attenuation relationship.
A Probabilistic Safety Assessment (PSA) procedure for Mixed Oxide (MOX) fuel fabrication facilities was developed. The procedure is a "two-part five-step" approach which takes characteristics of MOX fuel fabrication facilities into consideration. In the first part, so-called preliminary PSA, the hazard analysis approach was applied, which consists of two analysis steps: Functional Failure Modes and Effects Analysis (FFMEA) and Risk Matrix Analysis. The FFMEA analyzes a variety of functions of equipment composing the facility to identify potential abnormal events exhaustively. In the Risk Matrix Analysis, these potential events are screened to select abnormal events as candidates to be analyzed in the second part, using two-dimensional matrix based on the likelihood evaluated by probabilistic index method and maximum unmitigated radioactive release calculated by the Five-Factor Formula. For the selected abnormal events, in the second part, so-called detailed PSA, accident sequences, their occurrence frequencies and consequences are analyzed. These three analysis steps correspond to PSA procedure for nuclear power plant. The applicability of the PSA procedure was demonstrated through the trial application to model plant of MOX fuel fabrication facility.
The threat of terrorism is significantly increasing; thus the reinforcement of the scheme for physical protection of nuclear facilities is imperative in these days. To make the physical protection system for each facility really effective, it is necessary to establish a method to evaluate the performance of the system in comparison with a target set in advance. From this point of view, firstly, the history of development of the physical protection scheme is described; then, the status quo of physical protection system. Furthermore, in order to realize an effective reinforcement of the physical protection scheme for the facilities, the following viewpoints are provided in the paper; (1) Significance of discrimination between thefts of nuclear materials and sabotage of nuclear facilities; (2) Identification and discrimination of highly enriched uranium and plutonium in metallic state from other nuclear materials; (3) Necessity to reduce probabilities of success of theft to quasi-zero level from facilities which treat a certain amount of the above-mentioned materials; (4) For other facilities which do not treat such metallic nuclear materials, the target should be set mainly in view of the ability of prevention of sabotage.
As for the nuclear fuel reprocessing of the spent fuel, although there was argument of pros and cons, it was decided to start Rokkasho reprocessing project further at the Japan Atomic Energy Commission of "Long-Term Program forResearch, Development and Utilization of Nuclear Energy" in year 2004. The operation of Tokai Reprocessing is goingsteadily to reprocess spent fuel more than 1, 100 tons. In this paper, history, present status and future of reprocessing technology is discussed focusing from military Puproduction, Magnox fuel reprocessing to oxide fuel reprocessing. Amount of reprocessed fuel are estimated based on fueltype. Then, history of reprocessing, US, UK, France, Germany, Russian, Belgian and Japan is presented and compared ontechnology, national character, development organization, environmental protection, and high active waste vitrification.Technical requirements are increased from Pu production fuel, Magnox fuel and oxide fuel mainly because of higherburnup. Reprocessing technology is synthetic of engineering and accumulation of operational experience. The lessonslearned from the operational experience of the world will be helpful for establishment of nuclear fuel reprocessing technology in Japan.