[en] As an exchanging method of vacuum filter element installed at vacuum system of the Recycle Equipment Test Facility (RETF), the cask method is determined to use at a viewpoint of pollution expansion protection and radiation exposure reduction of workers. A principle proof test was conducted after trial protection of main parts according to design conducted in 1995 fiscal year. As a result, it was found that filter element (after storing a container) could be exhausted without any problem, in falling test of exhausting chute, that a setting method of exchanger onto upper part of the filter unit was required to improve, that a set of filter exchanging medium could be conducted scarcely any problem, that a load required to push a filter element into the determined position was at least 37 kg, and that an allowable interval at jointing with a double-door flange was 0.8 mm and air tightness could be kept if its inclination is less tha 0.85 mm. (G.K.)

[en] The purpose of the experiment is to establish a new nuclear fuel reprocessing system, which is aimed to achieve the extreme safety. In order to avoid any potential danger of explosion, all processes are made by the precipitation method at room temperature. The system consists of the following processes: 1. crystallization of uranyl nitrate from a dissolved fuel solution by cooling the solution; 2. complex formation of UO₂²⁺ and Pu⁴⁺ with carbonate ion by the addition of Na₂CO₃-NaHCO₃ solution adjusting pH to 9, followed by the separation of a precipitate containing the major fission products by the centrifugation method; 3. separation of Cs as a precipitate of cesium tetraphenylborate; 4. recovery of U and Pu as precipitates of hydroxo compounds from the alkaline solution by the addition of NaOH solution ; 5. separation of Sr from the precipitate in process 2; 6. recovery of NaHCO₃ from the NaOH solution by bubbling CO₂ gas. As a result, 99.95% of the U was recovered with the least amount of fission products. Pu are expected to be recovered in the same way as U. In conclusion, the present reprocessing system enables us to recover U and Pu in high ratios from spent nuclear fuel by means of a simple precipitation method, to separate hazardous Cs and Sr from high-level waste, and to exclude any potential danger owing to chemical processes under mild aqueous conditions. (author)

[en] The treatment of spent metal fuel from the EBR-II fast reactor commenced in June of 1996 at the Fuel Conditioning Facility on the Argonne-West site in Idaho, USA. During the first year of hot operations, 20 fuel assemblies entered processing and 6 low enrichment uranium product ingots were produced. Results are presented for the various process steps with decontamination factors achieved and equipment operational history reported

[en] As a part of the safety confirmation work of Tokai Reprocessing Plant, the appropriateness was checked on the basic data used in criticality safety and shielding design of early-designed facilities in the plant on the basis of recent knowledge and safety evaluation methods. In the criticality safety design, it was confirmed that critical and subcritical values concerning mass and concentration of U and Pu and equipment dimension were appropriate. In the shielding design, it was found that the relation between shielding thickness and permissible radioactivity might give underestimated results of shielding thickness necessary to limit dose rate to the designated one on some condition. In this cases, however, it was confirmed that necessary shielding thickness has been secured because of the conservative calculation conditions for the real conditions except the operation test laboratory (OTL). However, the amount of radioactivity handled at OTL needs to be limited. From a viewpoint of criticality safety, operational control for U and Pu transfer was also investigated. As a result of it, at the transfer route where erroneous batch-wise transfer of process solution might lead to a criticality accident, the reliability of U and Pu concentration measurement needs to be improved by multiple measurements. At other transfer routes, it was confirmed that single failure of equipment or operation error would not lead to a criticality problem. (author)

No abstract available

No abstract available

[en] This study concerns the separation of rare earths (RE) uranium, transuranium elements (TUE)from spent nuclear fuel solution by high performance liquid chromatography (HPLC). The correlation between retention behaviour and ionic radii of lanthanides and actinides are also discussed. The aim of this work is to develop a convenient separation procedure for RE, uranium and TUE by studying optimum physico-chemical conditions such as pH of the eluent, temperature and effect of the concentration of the matrix element. The experimental conditions were established by using a mixture of fourteen lanthanides plus yttrium and uranium of known concentration. The behaviour of plutonium was investigated by tracer experiment. Beta-gamma counting, alpha- and gamma-spectroscopy and coulometry were used for detection of the elements. The effect of pH shows that the retention time falls sharply, especially for lighter lanthanides, at lower pH (3.0-3.5) following a gradual decline in the pH range of 4.0-5.0, and then the retention time attains a plateau at high pH of 5.0-6.0. The separation features of RE, uranium and plutonium were satisfactory at 70-90 degree C, and 80 degree C was used as the working temperature. The effect of the concentration of matrix element, uranium, of nuclear fuel on the separation characteristics of RE was studied in the range of 70-700 μg (300-3000 nmol). No appreciable effect was observed up to 220 μg of uranium and above this concentration the elution profile of uranium overlaps with that of heavier lanthanides (up to Er). The elution profiles of spent nuclear fuel solution show that uranium elutes seven minutes earlier than that observed in the simulated experiment. Plutonium elutes at two different retention times (12 and 27 minutes), suggesting the presence of more than one species of plutonium in the spent nuclear fuel solution. Ruthenium elutes at 4 minutes of retention time. The elution profiles of americium and curium overlap, and europium elutes five minutes later than that observed in the simulated experiment. The retention and ionic radii profiles show the consistency for all of the studied lanthanides and actinides except for plutonium

[en] A slight change in the level-volume relation for an accountability tank for a large amount of plutonium nitrate solution was observed at the Plutonium Conversion Development Facility (PCDF) in the Power Reactor and Nuclear Fuel Development Corp. (PNC), Tokai Works. From the results of annual tank re-calibrations for the plutonium receiving tank from 1985 to 1992 using the incremental feed of nitric acid as the density standard, it became clear that the relation between the level and the volume changed slightly, and the rate of the change was a linear function of operating time. Also it became clear that the change was linear in relation to the level. In the PCDF, the cumulative change in the volume at the nominal level was evaluated to be 0.1% during 8 years' operation. It was also evaluated that the repeatability of the re-calibration is much better than 0.1%. A reasonable frequency of tank re-calibration is once every 5 years. (author)

[en] Short communication

[en] Rokkasho fuel reprocessing facility is the plant with maximum reprocessing capacity per year of 800 t-U, and the maximum storage capacity of spent fuel of 3000 t-U. One of its features from the viewpoint of criticality safety is the burnup of spent fuel assemblies was mortgaged, and made into the design condition for criticality safety. It is necessary to confirm the change of fuel composition accompanying burning in the case of this criticality safety design. In Rokkasho fuel reprocessing facility, it has been decided to be measured with the burnup measuring device. The flow of spent fuel through the reprocessing plant and the facilities of accepting and storing spent fuel are described. The method of criticality safety design for the racks of temporarily placing and storing spent fuel, the effect of burning considered in criticality safety analysis, and the measurement of average degree of residual enrichment of spent fuel assemblies with the burnup measuring device are reported. The burnup measuring device is the combination of the measuring methods for gamma-ray spectra, emitted neutrons and gamma-ray. The methods of measurement and analysis and the errors of the burnup measuring device are explained. (K.I.)