[en] Two comparative pulse irradiation experiments simulating reactivity initiated accidents (RIAs) were performed on high burnup PWR fuel rods under different coolant temperature conditions in the NSRR. The test RH 1 was carried out at the conventional NSRR test temperature of approximately 20.deg.C, while the coolant temperature was ∼280.deg.C in the test RH 2, corresponding to a hot standby temperature of PWR. In the both tests, the fuel rods did not fail against fuel enthalpy increases of 510 and 414J/g (122 and 99cal/g), respectively. The transient fuel behavior in the test RH 2, including cladding deformation and fission gas release (FGR), suggested that results from the high temperature test would follow the tendencies observed in previous tests at the room temperature, when the data is plotted as a function of peak fuel enthalpy. Numerical analysis using the RANNS code showed that the cladding deformation was driven only by the pellet cladding mechanical interaction (PCMI) and the following gas induced deformation did not occur in the test RH 2, because the rod internal pressure did not reach the threshold pressure

[en] MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO₂ dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)

[en] In order to evaluate a self-accelerated reaction in an evaporator in a fuel reprocessing plant due to organic-nitric acid reactions, a development of a calculation code is under way. Mock-up tests were performed to investigate the fluid dynamic behavior of the organic solvent in the evaporator. Based on these results, the model of the calculation code was constructed. This report describes the results of mock-up tests and the model of the calculation code. (author)

[en] Silver mediated electrochemical oxidation is considered as a promising candidate for the decontamination of TRU waste, due to its efficiency and safety. Fundamental experiments of the decontamination of solid waste and the destruction of organic liquid waste were carried out. This report summarizes the results of experiments. (author)

[en] This report presents outline of the nuclear fuel treatment facility for the purpose of preparing solution fuel used in Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) in Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF), including descriptions of process conditions and dimensions of major process equipments on dissolution system of oxide fuel, chemical adjustment system, purification system, acid recovery system, solution fuel storage system, and descriptions of safety design philosophy such as safety considerations of criticality, solvent fire, explosion of hydrogen and red-oil and so on. (author)

[en] Design and construction of a plutonium dissolver using silver mediated electrolytic oxidation method are promoted in NUCEF. Criticality safety analysis for the plutonium dissolver is described in this report. The electrolytic plutonium dissolver consists of connection pipes and three pots for MOX powder supply, circulation and electrolysis. The criticality control for the dissolver is made by geometrically safe shape with mass limitation. Monte Carlo code KENO-IV using MGCL-137 library based on ENDF/B-IV was used for the criticality safety analysis for the plutonium dissolver. Considering the required size for construction and criticality safety, diameter of pot and distance between two pots were determined. On this condition, the criticality safety analysis for the plutonium dissolver with connection pipes was carried out. As the result of the criticality safety analysis, an effective neutron multiplication factor keff of 0.91 was obtained and the criticality safety of the plutonium dissolver was confirmed on the basis of criteria of ≤0.95. (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] The silver mediated electrochemical oxidation (Ag/MEO) process with the ultrasound agitation has been developed for the purpose of the mineralization of organic wastes containing transuranium nuclides (TRU wastes) at the nuclear fuel reprocessing process. In the Ag/MEO process, organic solvents are decomposed by divalent silver cations under the relatively low temperature and the ambient pressure condition. The ultrasound agitation is effective in mixing the electrolytic solutions and the organic solvents, and is expected to promote the oxidation of the organic solvents. Therefore, the Ag/MEO process with the ultrasound agitation could be a candidate for the treatment of organic solvents. Destruction tests of tributylphosphate and dodecane by the Ag/MEO process were conducted to optimize some treatment conditions about electrolytic currents, solution temperatures, nitric acid concentrations and ultrasound intensities. Under optimized conditions, the destruction tests of kerosene and N,N,N',N'-tetraoctyl-3-oxapentane-1,5-diamide (TODGA) were carried out. It was confirmed that the Ag/MEO process is effective for the mineralization of these organic solvents. (author)

[en] It is important to decrease the radioactivity of transuranium (TRU) waste arising from reprocessing plants by the decontamination for its disposal. In order to dispose TRU waste safely and rationally, a decontamination technology is required to be developed. For this purpose, the Japan Atomic Energy Agency has conducted a basic study focusing on the cerium-mediated electrolytic oxidation (CeMEX) method. In this study, two series of tests were performed to confirm the sufficient corrosion rate for the decontamination of metallic waste with the CeMEX method. One is the pre-corrosion test to survey an optimum solution condition for the generation of cerium(IV) ion under different conditions in concentration of cerium(III) ion and nitric acid. The other is the corrosion test to evaluate the corrosion rate of stainless steel as simulating waste under the optimized solution condition. It was confirmed that the average corrosion rate of stainless steel was 3.3 μm/h for 90 hours. This means that the decontamination can be completed within 6 hours and that the decontamination solution can be recycled 15 times, assuming that the decontamination to the clearance-level needs corrosion depth of 20 μm. From the results, the CeMEX method is sufficiently applicable to the decontamination of TRU waste. (author)

[en] In order to clarify the failure mechanism and determine the failure limit of the High-Temperature Gas-cooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 0.578 to 1.869 kJ/gUO₂ in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above 1.40 kJ/gUO₂, where the peak fuel temperature reached over the melting point of UO₂ fuel. It was concluded that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions. (author)