[en] Recently, many U.S. nuclear power stations need the on-site interim spent fuel storage installations because of the nuclear fuel cycle policy of the government that the spent fuel is not reprocessed. The NUHOMS system developed by NUTECH is rapidly spreading for the dry spent fuel storage by its good economics in the U.S.A. This report describes the unique NUHOMS (NUTECH Horizontal Modular Storage system) made of reinforced concrete that is not so familiar in Japan like U.S.A. (author)

[en] Designing a reactor shield, radiation heat and its removal are prime considerations. Calculation of the radiation heating and heat removal requires many steps and troublesome procedures. Therefore, the RATEX code system has been developed for a systematic calculation of radiation attenuation, radiation heating and temperature distribution. To eamine the code system, it was used for radiation heating and temperature analysis at the shield system. Availability of the code system was successfully proved. (author)

[en] Analysis of graphite oxidation in a versatile high temperature gas-cooled reactor (VHTR) at the time of the accident of air ingress after the breakdown of a primary piping was performed. An equation for the transport of material in the channel and an equation for the diffusion of oxygen in graphite were used, and the computer code for the analysis was developed. It was applied for the present situation. The air entered into the channel of coolant transfers on the surface of graphite, diffuses into the graphite and oxidizes the graphite. The temperature distribution in the reactor and the natural circulating flow were obtained by a two-dimensional unsteady heat transfer code 'KATEC'. The distribution of burn-off weight loss in the depth direction of graphite was estimated by the analysis code of graphite oxidation 'GRABOC' for calculating the transfer of oxygen to graphite surface and the diffusion of oxygen in the graphite. As the results of analysis, it was found that the strength of graphites in the lower plenum post near air entrance and the fuel region was reduced remarkably. (Kato, T.)

[en] In the case of a large rupture accident in the primary coolant system of VHTR, it will be necessary to suppose the long term graphite oxidation in the reactor due to the air ingress with natural convection. Therefore, the accident analysis code for the evaluation of graphite oxidation accident has been developed. The results of the analysis on the basis of VHTR conditions show that the portion of graphite with higher temperature and with higher oxygen partial pressure has the larger loss of the surface, and that the portion of graphite with lower temperature and with higher oxygen partial pressure has the deeper burn-off effect. (author)

[en] In an experimental multi-purpose high temperature gas reactor (VHTR) being developed in Japan, because of the high temperature fuel, the quantity of fission product released from the coated fuel particles is considerable, which then deposits in the primary cooling system as plate-out. This plate-out phenomena were calculated by the use of the code PADLOC for the calculation of the distribution of fission-product deposition. The values were compared with those obtained with the in-pile gas loop OGL-1 installed to the JMTR (Japan Material Testing Reactor). Using the code PADLOC which assumes the material transfer in flow-paths and surface-layer adsorption equilibrium, the tendency of decrease in the concentration distribution from upstream to downstream was enhanced in comparison with that measured in the OGL-1. In this connection, a microscopic model of decreasing the material transfer rate was presented, so that the results by the calculation with the code PADLOC agreed with those measured in the gas loop OGL-1. (Mori, K.)

[en] This report presents the results of quantitative evaluation on the effects of the dominant parameters on the maximum fuel temperature in the core thermal hydraulic design of the High Temperature Engineering Test Reactor(HTTR) of 30 MW in thermal power, 950 deg C in reactor outlet coolant temperature and 40 kg/cm² G in coolant pressure. The dominant parameters investigated are 1) Gap conductance. 2) Effect of eccertricity of fuel compacts in graphite sleeve. 3) Effect of spacer ribs on heat transfer coefficients. 4) Contact probability of fuel compact and graphite sleeve. 5) Validity of uniform radial power density in the fuel compacts. 6) Effect of impurity gas on gap conductance. 7) Effect of FP gas on gap conductance. The effects of these items on the maximum fuel temperature were quantitalively identified as hot spot factors. A probability of the appearance of the maximum fuel temperature was also evaluated in this report. (author)

[en] The conceptual design study of 1000 MWe gas-cooled fast breeder reactor (GCFR), which is used in the GCFR-VHTR symbiotic energy system, has been performed. In this report, the transient response of the GCFR core to accident events has been analyzed and safety performance of the 1000 MWe GCFR has been evaluated considering the analyses. A depressurization accident caused by failure of a primary coolant system and a reactivity insertion accident due to withdrawal of a control rod have been analyzed using nuclear and thermo-hydraulic coupled program MR-X developed for kinetics analysis of gas-cooled fast breeder reactors. The maximum fuel and cladding temperatures are most important problem to be analysed during a trangient of a gas-cooled fast breeder reactors. The analyses show that reliable reactor shutdown and emergency cooling systems are most important to achieve successful cold shutdown well before leading to core damage and also that no severe failures of fuel pin and cladding occures by working above mentioned safety systems well during the accidents. (author)