[en] In Japan, among these one to two years interest on safety problem of the nuclear fuel cycle facility has increased, and importance on establishment of safety technique basement on reprocessing process and plutonium safe treatment has also strengthened. Under such states, a large change is appearing with the nuclear fuel cycle such as promotion of plutonium application to light water reactor, investigation start on concretization of intermediate storage of used fuels except power plant, re-investigation of breeding reactor developing plan, and others. Now, Japan Atomic Energy Research Institute intends to successively and strongly promote study on the nuclear fuel cycle safety shown as follows under intimate cooperation with some research institutes in and out of Japan: 1) Reinforcement of safety technique basement for supporting safe operation of the Rokkasho reprocessing plant, 2) Study on response to promotion of plutonium application to the light water reactor, and 3) Planning on future reprocessing technique and others. (G.K.)

[en] A computer simulation model has been developed to analyze the basic characteristics of heat-exchanger type steam-methane reformer which is the key component to produce hydrogen using the nuclear process heat from high temperature gas cooled reactor. This model is based on the one-dimensional one taking account of heat transfer and reaction kinetics. This report describes the analytical model, the solution procedure and the calculation results on gas temperatures, reaction rates, chemical equilibrium attainments and heat flux along reformer tube. (author)

[en] We prepared 150kgU of 10%²³⁵U uranium solution for the critical assemblies (STACY, TRACY) by dissolution of mixture 1.5%²³⁵U uranium dioxide pellets and 12%²³⁵U uranium dioxide pellets with the fuel treatment system of NUCEF. In order to find optimum operation conditions for dissolution, we carried out preliminary experiment using each one pellet and characteristic experiment using dissolver. In this report, results of these experiments and operation were described. As a result of these experiments, we obtained following operation conditions; initial nitric acid 7M, temperature 80degC, operation time 8h. Under these conditions, we dissolved UO₂ of over 99% satisfactorily and prepared 150kgU of fuel solution. (author)

[en] At nuclear fuel cycle safety engineering research facility(NUCEF), aiming at the securing of safety, the advance of technology and the perfection of technical base in the back end of nuclear fuel cycle, it was decided to carry out the research and development on criticality safety, advanced fuel reprocessing process, TRU waste management, TRU chemistry and the elementary technology related to the NUCEF. The experimental study on the criticality characteristics of nuclear fuel, the experimental study on transient criticality events, the research on the techniques of assessing criticality safety, the test for verifying the safety of reprocessing facilities, the research on advanced reprocessing process, the research and development of TRU waste treatment and disposal, the research on the quality inspection method for TRU waste solidified bodies and the measurement technology, the test of spent fuel characteristics, the basic research on reprocessing based on new principle, the solid chemistry research on TRU, the development of safety simulation technology, the development of maintenance and inspection technologies and so on are planned. (K.I.)

[en] The bearing loads were measured and analyzed for the gas bearing type circulator of the Helium loop HENDEL. From results of the study, it was found that the static load acting to bearing pads in the journal bearing does not remain constant as expected and is not isogonal due to the external forces from momentum changes and pressure gradient of the working gas and from electromagnetic force in the driving motor. As to the dynamic load in the bearing, it was found to depend on the static load and was also found that the absolute value of its vector is not constant in a rotation of the shaft. It was predicted and reasoned that the divergent shaft vibration is caused by the vibrational dynamic load in absolute value and relatively low stiffness of the spring-pivot. To improve weak points of the gas bearing circulator, some new design criteria are offered and the substantial methods are also indicated to realize those criteria. (author)

[en] NUCEF consists of two criticality experiment devices, three alpha-gamma cells, the experimental equipments contained in about 90 glove boxes and many others. The background and the course of the construction and fitting of NUCEF are described on the selection of important research, the rationalization of facilities, the integration of the criticality safety experiments facility(CSEF) and the TRU safety test facility(TRUST), the safety examination, and the construction and fitting. The whole NUCEF, the static criticality facility(STACY), the transient criticality facility(TRACY), the mock-up test of the machinery and equipment for both criticality facilities, the solution fuel preparation facility, cells and glove boxes and the equipment contained in them, the analysis facility consists of four analysis rooms, and the radioactive waste facility for gas, liquid and solid wastes are described. The classification of these facilities based on the relevant laws is shown. The way of thinking in the basic design and the detailed design is explained. The permission of nuclear reactor installation and the use of nuclear fuel substances was obtained. The design and the method of works were approved. The inspection is reported. The measures for the security are explained. (K.I.)

[en] Experimental studies on heat transfer and flow characteristics of a simulated fuel rod for the VHTR (Very High-temperature Gas-cooled Reactor) has been performed with the fuel stack test section (T₁) of the Helium Engineering Demonstration Loop (HENDEL), using a helium gas of almost same conditions of the VHTR operation. This report describes test result obtained by single-channel test rig of T₁. Test conditions are as follows; Inlet temperature : 290 -- 620 K, Inlet pressure : 0.4 -- 4.0 MPa, Inlet Reynolds number : 1,600 -- 21,000, A simulated fuel rod, Electrical input : Maximum 90 kW, Heat flux distribution : Uniform. The conclusions derived from the tests are that friction factors and heat transfer coefficients are about 20 %, 15 -- 60 % higher than those for concentric smooth annuli, respectively. The reason may be due to the effect of spacerribs. (author)

[en] A series of critical experiments with 10% enriched uranyl nitrate solution using a cylindrical core tank 60 cm in diameter have been performed with the Static Experiment Critical Facility at the Nuclear Fuel Cycle Safety Engineering Research Facility in the Tokai research establishment of the Japan Atomic Energy Research Institute. In the first series of experiments using the cylindrical core tank, systematic data of the critical height for water-reflected cores and unreflected cores were obtained by changing the uranium concentration of the fuel solution from 313 to 225 g U/ell. As the reactivity of each core is controlled only by solution height, these criticality configurations, which have simple cylindrical shapes, are available for the validation of calculation codes used in criticality safety designs of nuclear fuel cycle facilities. The neutron multiplication factors of experimental cores were calculated with the two-dimensional transport code TWOTRAN in the SRAC code system and with the continuous-energy Monte Carlo code MCNP4A, employing the Japanese evaluated nuclear data library JENDL-3.2. The calculations from the combination of these calculation codes and the nuclear data library reproduce the neutron multiplication factors within an error of 0.9% for the experimental configuration of critical cores

[en] The helium engineering demonstration loop (HENDEL) at JAERI is designed as a large-scale model test facility for demonstrative operation of high-temperature components, such as a fuel stack, an in-core structure, an intermidiate heat exchanger, high-temperature pipings and valves of the experimental very high-temperature gas cooled reactor (VHTR). The HENDEL is supposed to be operated under the simulated conditions of the VHTR. The HENDEL consists of the first and second helium gas loops (M₁ and M₂ loops) and six test sections, the first of which was completed and the others are being designed. This report describes operational data of components (heater, blower and cooler) and overall performance of the first loop (M₁ loop), during the test operations until March, 1983. M₁ loop is to provide helium gas of 450 ⁰C, 0.4 kg/s and 4.0 MPa to a fuel stack test section (T₁). (author)

[en] A fuel block composed of hexagonal prism-shaped graphite block and fuel rods is under development for the experimental very high-temperature gas-cooled reactor (VHTR) at the Japan Atomic Energy Research Institute (JAERI). Helium gas flows in an annular channel between a fuel hole and a fuel rod. In order to investigate the heat transfer and flow characteristics in the channels of the fuel block, the simulated fuel block test section called multi-channel test rig was set in the Helium Engineering Demonstration Loop (HENDEL) of JAERI. The present paper reports the experimental results obtained under the conditions of uniform heating of 12 fuel rods. It was found that the He gas flow rate was almost uniform in all of the fuel channels and that the temperature distribution in the horizontal cross section of the fuel block was also uniform. As for the heat transfer, the measured heat transfer coefficient was higher than the values predicted with heat transfer correlations of a smooth annular channel used as a design basis for VHTR. (author)