[en] Purpose: To easily gasify spent contaminated moderators and the like of a graphite-moderated reactor with CO₂. Method: Graphite is dipped in an aqueous solution of nitrate of a gasifying catalyst using as a solvent nitric acid. For example, when a graphite block or compact is dipped in a catalyst solution consisting of 1 to 14M of HNO₃ and 0.22M of Fe(NO₃)₃ at a temperature in the range of from 20 to 90⁰C for one to five hours, 400 to 2000 ppm of iron is added up to the central part of graphite. After graphite has been dried, 100% of CO or N₂ is caused to flow into the solution at 850 to 1000⁰C for about 30 minutes to activate the catalyst. As a consequence, the gasifying catalyst is uniformly added to graphite, whereby it is possible to increase the gasification velocity of graphite with CO₂ at a temperature in the range of 850 to 1000⁰C. (Nakamura, S.)

[en] Critical control can be achieved and decontamination factor increased by adding metallic ions having a large neutron absorption cross section, such as Hf, Sc and In, to the dissolution step and/or the extraction and separation steps of a spent nuclear fuel reprocessing method. This method performs critical control by the addition of soluble neutron poison metallic ions, and restrains the decrease of the decontamination factor of fission products and the reduction of recovery of plutonium and/or super plutonium by masking acidic organic phosphorus compounds. As the result of experiments, it was found that the increase in distribution ratio of radioactive zirconium due to formation of dibutyl phosphate can be restrained by the addition of a variety of metallic ions. A nitric acid solution containing U, Pu and fission products were contacted with tributyl phosphate/dodecane solution. The extracted U and Pu were washed with nitric acid solution. After 7 to 8 hours of operation, the compositions of extraction waste and extracted media were analyzed. The plutonium recovery was 99% or more. The material balance was 97-99%. The distribution and the concentration were satisfactory. The addition of nonradioactive zirconium increased by four times the decontamination factor once lowered to 1/8. (Iwakiri, K.)

[en] OECD (Organization for Economic Cooperation and Development)/NEA (Nuclear Energy Agency) held a workshop 'A Part of Research in Regulation' in Paris in July, 2001. One of recommendation requested a proposal on 'a plan of regulation and industry co-operation on safety research with each independency by investigating and analyzing merits and demerits'. Therefore, GRIC (Group on Regulation and Industry Co-operation on safety research) organized by specialists was established and begun investigation in November, 2001. The report: OECD/NEA 'Regulator and Industry Co-operation on Safety Research: Challenges and Opportunities' (2003) was opened as a results. In this paper, abstract of the report and the future issues in Japan was stated. An outline of GRIC meetings, a participating organization, the state of cooperation in each country, merits and demerits, the prospects, the proposal and future issues in Japan are reported. (S.Y.)

[en] With the view of continuous analysis of UF₆ and/or F₂ in the fluorination of U for the development of the fluoride volatility process, an in-line gas chromatograph, made of anti-corrosive materials against the fluorides, has been constructed using two different columns and automatic sampling system. The separation columns used, were optimized for the simulated process gas composition ((N₂, O₂, F₂)/(UF₆)=18, (N₂, O₂)/(F₂)=3.3), in order to obtain the complete peak separation. The column selected for UF₆ analysis is 0.4 cm phi Cu tubing of 3 m long packed with 40w/0 of poly-trifluoromonochloro-ethylene oil which is loaded on poly-tetrafluoro-ethylene powder. The column for F₂ analysis is the combination of the above column with a conversion column of 0.4 cm phi tubing of 2 m long packed with KCl powder. Under these column conditions, the analytical time is 9 and 3 min for UF₆ and F₂, respectively at the flow rate of 100 ml/min. The calibration curves are linear with a zero intercept, and the analytical limits are, respectively 0.2 and 1 mmHg in partial pressures of UF₆ and F₂ with 10 ml of sampling volume. In actual adaptation to the fluorination of uranium oxides, the in-line gas chromatograph was satisfactorily able to measure the reaction rate and to detect the end of fluorination. (auth.)

[en] With the operation of a fuel reprocessing plant in the Power Reactor and Nuclear Fuel Development Corporation (PNC) and the plan for a second fuel reprocessing plant, the research on fuel reprocessing safety, along with the reprocessing technology itself, has become increasingly important. As compared with the case of LWR power plants, the safety research in this field still lags behind. In the safety of fuel reprocessing, there are the aspects of keeping radiation exposure as low as possible in both personnel and local people, the high reliability of the plant operation and the securing of public safety in accidents. Safety research is then required to establish the safety standards and to raise the rate of plant operation associated with safety. The following matters are described: basic ideas for the safety design, safety features in fuel reprocessing, safety guideline and standards, and safety research for fuel reprocessing. (J.P.N.)

[en] Recent trends and problems in the application of dry techniques to the aqueous reprocessing system are generally reviewed. The first part of this report explains the special features and merits of various dry reprocessing techniques referring to a summary table. The second part reviews the technological development and problems of dry techniques since about 1970. The fluoride volatilizing and the pyrometallurgical processes are especially taken up here mainly from the viewpoint of the fuel reprocessing for fast reactors. The third part discusses the adaptation of dry techniques to the pretreatment and after-treatment processes in the wet reprocessing system. As for the pretreatment processes of fast reactor fuel, the high-temperature clad-removing with the pyrometallurgical process, the voloxidation method, and the treatment of solvent hull and off-gas are discussed. Some discussions are also made on the pretreatment process for the reprocessing of high temperature gas reactor fuel. As for the after-treatment processes, the solidification of waste liquids and the conversion process are briefly discussed. (Aoki, K.)

[en] Nuclear Material Control Center has been studying inventory estimation method for an evaporation concentrator installed in large scale reprocessing plant. Solution volume and nuclear material concentration are required to estimate the inventory. Since the evaporation concentrator will be operated continuously, drawing solution out of the evaporation concentrator is impossible. Therefore we made out two equations to estimate solution volume and nuclear material concentration using instrumental signals (e.g. level, density, temperature). One correlates solution volume to the flow rate of feed solution and heating steam. The other correlates solution concentration to density and temperature. It was recognized that reproducibility of each equations are less than 0.5% and 1.5%. (author)

[en] In order to advance the nuclear fuel recycling in Japan, the safety in the back end of fuel cycle must be secured and the advance of technology toward 21st century is necessary. For the purpose, Japan Atomic Energy Research Institute is constructing the nuclear fuel cycle safety engineering research facility (NUCEF), and at present, the general function test is carried out, aiming at the hot test in fiscal year 1994. The viewpoint in the research and development related to the back end of nuclear fuel cycle and the main subjects and the research plan at NUCEF are shown. The aim of the NUCEF project is to secure the safety, to advance the technology and to perfect the technical base. The perspective of the NUCEF project is given. (K.I.)

[en] Separation and decontamination factors for three steps of an extraction process, i.e., extraction, scrubbing and stripping, as well as an overall decontamination factor for the extraction process as a whole are defined to show the separation efficiency in the process. The equations are derived between these factors and experimental values such as distribution ratios, scrubbing percents and stripping percents. The equations are applied to the separation of fission products/uranium with tributyl phosphate (TBP) and seven alkyl amines irradiated to the extent of 10⁸ r, in order to follow the radiation effect on separation and decontamination factors. Decontamination factors decrease with increasing dosage in the TBP system, whereas they increase in most amine systems. N-cyclohexyl dilauryl amine and TBP are the best for the decontamination among unirradiated extractants. Amberlite LA-2 is the best among irradiated. In most cases a scrubbing step is more important than an extraction step as far as the decontamination of fission products/uranium concerns. (author)

[en] The reprocessing of the spent fuel which is taken out from nuclear reactors is very effective and beneficial from the viewpoint of resources. This reprocessing consists of the removal of fission products and the extraction of uranium and plutonium from spent fuel. The positioning of reprocessing in the nuclear fuel cycle, the composition of the spent fuels in light water reactors, fast breeder reactors and high temperature gas cooled reactors especially for the variable burn-up and cooling time, the decay heat of the fission products in spent fuel, the economic value of spent fuel, the amount of spent fuel coming out of light water reactors, fast breeder reactors and high temperature gas cooled reactors with the capacity of 1000 MWe each, the special features of reprocessing, the outline of reprocessing methods, for example, the purex process, the thorex process, the halide volatilization process, the high temperature volatilization process and the pyro-metallurgical chemical process, the technical comparison between these processes, the technical problems for these processes, the detail of these processes and their technical characteristics, the facilities for each process, the off gas in reprocessing plants, the radioactive waste disposal in each process and the relating research and development are explained in this paper. (Nakai, Y.)