[en] Chemical and chemical engineering techniques of the uranium ore processing established by France COGEMA (Compagnie Generale des Matieres Nucleaires) have been comprehensively reviewed in preparation for successful test operation of the pilot plant to be completed by the end of 1981. It was found that the amount of sulfuric acid (75 Kg/t, ore) and sodium chlorate (2.5 Kg/t, ore) recommended by COGEMA should be increased up to 100 Kg/t, ore and 10 Kg/t, ore respectively to obtain satisfactory leach of uranium for some ore samples produced at the different pits of Goesan uranium mine. Conditions of the other processes such as solvent extraction, stripping, and precipitation of yellow cake were generally agreed with the results of intensive studies done by this laboratory

[en] The system analysis of an advanced spent fuel management process to establish a non-proliferation model for the long-term spent fuel management is performed by comparing the several dry processes, such as a salt transport process, a lithium process, the IFR process developed in America, and DDP developed in Russia. In our system analysis, the non-proliferation concept is focused on the separation factor between uranium and plutonium and decontamination factors of products in each process, and the non-proliferation model for the long-term spent fuel management has finally been introduced. (Author). 29 refs., 17 tabs., 12 figs

[en] Short communication

[en] The Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical dry reprocessing technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. With this technology, a significant reduction of the volume and heat load of spent fuel is expected, which could decrease the burden of safety and economics. In this study, MCNPX code calculations were carried out to estimate the performance of a neutron coincidence counter designed for measruement of the process materials in the pilot-scale ACP facility. To verify the design requirement, the singles and doubles counting rates of the detectors were simulated with the latest coincidence capability of the MCNPX code. Then, the precision of the coincidence measurements were evaluated on various process materials from the ACP. It was verified that the performance of the neutron coincidence counter could meet the design criteria for all samples in the ACP, and the material accounting system for the pilot-scale ACP facility could meet the IAEA safeguards goals.

[en] At Korea Atomic Energy Research Institute (KAERI), we investigated the corrosion behavior of a series of Fe-Cr-Ni alloys with different chromium contents in molten LiCl and molten LiCl-25wt%Li₂O mixture at temperatures ranging from 923 to 1123 K. In molten LiCl, dense protective scale of LiCrO₂ grows outwardly while corrosion is accelerated by addition of Li₂O to LiCl. The basic fluxing of Cr₂O₃ by Li₂O would be the cause of accelerated corrosion. Because of low oxygen solubility and very high Li₂O activity in the molten LiCl-Li₂O mixture, Cr is preferentially corroded while Ni remains stable and thus, corrosion rate of the alloys in molten LiCl-Li₂O mixture increases with an increase in Cr content

No abstract available

[en] Full text: Nuclear energy plays a major role of supplying stable and sustainable energy in Korea. Currently, 18 nuclear power units with 15,716 MWe capacities are in operation, and the nuclear share in power generation is about 40 percent. As the utilization of nuclear energy increases, the management of the spent fuel becomes an imminent task to be resolved. While the ultimate policy of the back end fuel cycle is still under consideration in Korea, the main direction of the research is to develop the proliferation resistant fuel cycle technology to reuse the spent fuel for effective management of the accumulated spent fuel in a proliferation resistant way. Since Korea is developing the improved nuclear reactor system utilizing the oxide fuel for the near term deployment, and innovative nuclear system utilizing the metallic fuel for the long term application, the proliferation resistant fuel cycle technology development has mainly two directions of research: DUPIC (Direct use of spent PWR fuel in CANDU reactors) technology for the oxide fuel system and the pyroprocessing technology for the metallic fuel system, respectively. There has been a significant achievement in the DUPIC technology development since its beginning in 1991. The basic concept of the DUPIC fuel cycle is to directly fabricate the CANDU fuel from the spent PWR fuel by using thermal/mechanical processes in hot cells without the separation of fission products and transuranic elements. KAERI has successfully fabricated several DUPIC fuel elements at hot cells remotely, and the performance evaluation through a series of irradiation tests at the Hanaro research reactor is underway. DUPIC technology is developed by international joint research with Canada, the USA, with participation of the IAEA, and is internationally acknowledged as a typical proliferation resistant fuel cycle technology. The current status of the development and the requirements for practical application will be discussed in the presentation. The development for the pyroprocessing technology in Korea has recently been embarked upon. The main objectives of the research are to develop the electrolytic reduction method of the spent oxide fuel, and to fabricate the metallic fuel for use in the future innovative nuclear system of fast neutron spectrum. It will be performed in close relations with the international Generation IV research program, and the basic research using inactive surrogate material is currently underway. The current status and future plan of the development will be discussed in the presentation. As both fuel cycle technologies are well recognized from the viewpoint of the proliferation resistance aspect, the requirements of the future fuel cycle technology for the enhanced proliferation resistance with improved performance and economy will be discussed. (author)

[en] Full text of publication follows: KAERI is developing technologies which could reduce the increasing amount of spent fuel and dramatically decrease the disposal load, through recycling a waste salt in a pyro-processing system. This study aims at providing a new way to minimise the waste salt to be disposed of, while removing the fission products generated during a pyro-processing procedure. The main pyro-processing processes being developed by KAERI include a voloxidation for pulverising an oxide fuel pellet to an oxide powder, an electroreduction for converting an oxide powder into a metal by using LiCl salt, an electro-refining for recovering uranium from the converted metal by using a LiCl-KCl eutectic mixture and a waste treatment for treating the fission products arising from these processes. KAERI is presently performing a study on the removal of Cs, Sr and lanthanides including Y from a pyro-process. First, KAERI has been testing various technologies for a partitioning of Cs and Sr from a spent nuclear oxide fuel. Among them, a typical method is to remove Cs in the form of an oxide gas from the voloxidation process and then to remove the remaining Sr in a precipitate form by adding an inorganic such as carbonate to the waste LiCl salt from the electroreduction process. Another way is to simultaneously partition both Cs and Sr from a waste LiCl salt by using either Czochralski or zone freezing technologies. Finally, rare-earth elements including Y in the spent LiCl-KCl waste generated during the course of an electrorefining are removed in the form of an oxide precipitate by using an air oxidation. KAERI is focusing on a total recycling of the waste salts to each process unit after removing the fission products arising from different salts, with a minimum waste salt release to a permanent repository. This paper provides the experimental conditions to separate these fission products from each process unit and also evaluates the separation efficiency. In addition, this work will present and discuss the waste forms to be disposed of and a reference flow sheet for removing the fission products from a pyro-process. (authors)

[en] Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

[en] In preparation for the test operation of the pilot plant recently constructed with the capacity of 3 tons ore per day in uranium ore processing, some technical research and preliminary studies were carried out. In the study of solid-liquid separation, settling and filtering tests on slurries before and after leaching were carried out for the establishment of optimum flocculating condition compatible with the properties of slurry. From the data of settling velocity and final pulp density, thickener unit areas were also calculated. In the study of continuous operation of leaching,solvent extraction process, laboratory scale continuous type reactors such as leaching tanks and mixer-settlers were constructed and tested prior to the test operation of pilot plant. The design parameters being used were appropriate for the successful operation and the results of continuous running for efficiencies were very close to those of batch tests. Organic solvent loss into aqueous phase by its solubility was also studied. (Author)