[en] Kinetics and mechanism of radiolysis-induced free-radical reactions of xanthene and thioxanthene in halocarbons were studied. Detailed kinetic measurements performed in the temperature range of 0-65 deg. C allowed determinations of the reaction rate constants involving 9-xanthenyl radicals. The activation energy was estimated as 14 and 42 kJ mol^-1 for dissociation of xanthene/Cl π-complexes and peroxyl radicals, respectively. The spectral evidence of the intermediacy of thioxanthene/Cl π-complexes in the formation of the 9-thioxanthenyl radical in carbon tetrachloride was obtained. Using a combined pulse radiolysis-laser flash photolysis method, photochemistry of the 9-thioxanthenyl radical was studied. The excited 9-xanthenyl radical undergoes chlorine atom abstraction from the solvent with quantum yields of 0.10 and 0.30 in 1,2-dichloroethane and carbon tetrachloride, respectively. (author)

[en] The photochemistry of 9-xanthenyl radicals produced by pulse radiolysis of xanthene in halocarbon was studied by means of a successive laser flash photolysis in the presence and absence of oxygen. In deaerated solutions rapid (during 6 ns laser pulse) and permanent photobleaching due to chlorine atom transfer from solvents to the excited 9-xanthenyl radical was observed with quantum yields of 0.04 and 0.26 in 1,2-dichloroethane and CCl₄, respectively. In the solutions containing oxygen, equilibrium between 9-xanthenyl radicals and peroxyl radicals was established and recovery of the photobleached 9-xanthenyl radicals was observed, which was accounted for by dissociation of peroxyl radicals. The whole reaction scheme of formation and decay of 9-xanthenyl radicals in CCl₄ is discussed based on the kinetic simulations. (author)

[en] The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA, with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950degC (Saito, 1994). Test researches are being conducted using the HTTR to improve HTGR technologies and to collaborate with domestic industries to contribute to foreign projects for acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are being developed using data obtained with the HTTR, which include reactor kinetics, thermal-hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). A three-gas-circulators trip test and a vessel-cooling-system stop test were planned as a loss-of-forced-cooling test and demonstrate the inherent safety features of HTGR. The vessel-cooling-system stop test consists of stopping the vessel-cooling-system located outside the reactor pressure vessel (RPV), to remove the residual heat of the reactor core as soon as the three-gas-circulators are tripped. All three-gas-circulators is tripped at 9 MW. The primary coolant flow rate is reduced from the rated 45 t/h to 0 t/h. The control rods are not inserted into the core and the reactor power control system does not operated. A core dynamics analysis of the loss-of-forced-cooling test of the HTTR is performed. Analytical results for the reactor transient during the test are presented in this report. It is determined that the reactor power immediately decreases to the decay heat level due to the negative reactivity feedback effect of the core, even though the reactor shutdown system is not operational, and that the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Furthermore, the relation between the reactivities (namely, the Doppler, moderator temperature, and xenon reactivities) affecting re-critical time and reactor peak power level and total reactivity is addressed. The analytical results will be utilized for the design and construction of the Kazakhstan High Temperature Reactor (KHTR) and the realization of commercial Very High Temperature Reactor (VHTR) systems. (author)

[en] Enhancement of safety and cooling performance of fuel elements are desired for a commercial High Temperature Gas-cooled Reactor (HTGR). Applying sleeveless fuel elements and dual side directly cooling structures with oxidation resistant SiC-matrix fuel compact has a possibility of improving safety and cooling performance at the pin-in-block type HTGR. The irradiated effective thermal conductivity of a fuel compact is an important physical property for core thermal design of the pin-in-block type HTGR. In order to discuss the irradiated effective thermal conductivity of the SiC-matrix fuel compact which could improve the cooling performance of the reactor, the maximum fuel temperature during normal operation of the pin-in-block type HTGR with dual side directly cooling structures are analytically evaluated. From these results, the desired irradiated thermal conductivity of SiC matrix are discussed. In addition, the suitable fabrication method of SiC-matrix fuel compact is examined from viewpoints of the sintering temperature, the purity and the mass productivity. (author)

[en] JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950degC. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on postirradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/thm. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results. (author)

[en] FORNAX-A is a calculation code for amount of fission product (FP) released from cylindrical fuel elements of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation how to change FORNAX-A for enhancement of type of used diffusion coefficients, application to spherical fuel elements and calculation taking into consideration of temperature distribution of coated fuel particles in fuel element system. (author)

[en] We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)

[en] Fuel for the Very High Temperature Reactor (VHTR) is required to be used under severer irradiation conditions and higher operational reactor temperatures than those of present high temperature gas cooled reactors. Japan Atomic Energy Agency has developed the advanced silicon carbide (SiC)-coated fuel particles having thicker layer thicknesses, and zirconium carbide (ZrC)-coated particles that are expected to preserve their integrity at higher temperatures and burnup conditions than current conventional coated fuel particles. These particles have been fabricated successfully in order to perform irradiation tests at experimental reactors. This paper is summarized fabrication data of irradiation samples. (author)

[en] Primary helium circulators of the HTTR are the important components to circulate primary coolant of helium gas. Three circulators for the primary pressurized water cooler and one for the intermediate heat exchanger are installed in primary cooling system. The filter has been installed in upper casing of helium circulator to prevent that the fine particles in helium gas enters the gas bearing of circulators. The differential pressure of this filter rose gradually during rise-to-power tests. The rise of the filter differential pressure of the helium circulator causes the problem for reactor operation. Therefore, the filters were newly manufactured, and replacement of the filter was carried out. In replacement of the filter, appearance confirmation was carried out and attached substances were analyzed. This paper describes replacement work of the filter and investigation of cause of filter differential pressure rise. (author)

[en] In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically. (author)