• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.175-183
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• 2004

For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada deuterium uranium (CANDU) reactor, direct use of spent pressurized water reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-though and DUPIC fuel cycles.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.665-674
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• 2013

This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.997-1001
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• 2016

Conductive ceramics are being developed to replace current Pt anodes in the electrolytic reduction of spent oxide fuels in pyroprocessing. While several conductive ceramics have shown promising electrochemical properties in small-scale experiments, their long-term stabilities have not yet been investigated. In this study, the chemical stability of conductive $La_{0.33}Sr_{0.67}MnO_3$ in $LiCl-Li_2O$ molten salt at $650^{\circ}C$ was investigated to examine its feasibility as an anode material. Dissolution of Sr at the anode surface led to structural collapse, thereby indicating that the lifetime of the $La_{0.33}Sr_{0.67}MnO_3$ anode is limited. The dissolution rate of Sr is likely to be influenced by the local environment around Sr in the perovskite framework.

• Journal of Advanced Engineering and Technology
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• v.4 no.4
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• pp.431-436
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• 2011

This study was carried out to reduce the problem during distillation process, which separate U, TRU (TRans Uranium) metal electro deposit, Cd and LiCl-KCl eutectic salt generating from LCC (Liquid Cadmium Cathode) electro winning process. The cadmium recovering apparatus was manufactured to separate for each metal using solid-liquid separation method. The apparatus consists of the first sieve for the separation of U and TRU metal electrodeposit, the second sieve for the separation of LiCl-KCl eutectic salt, cadmium collection basket, and a heating furnace. In addition, the size of each sieve is 2 mm to 3 mm. In this experiment, a metal wire was employed to replace TRU metal electrodeposit and U, which exist actually in a LCC crucible. In the solid state, The LiCl-KCl is separated at 340℃ at which the solid and the liquid of the remaining cadmium and LiCl-KCl eutectic salt coexists in each, after the metal wire separated at 500℃. As a result, it seems that it would be beneficial to set the processing condition in the distillation process with the additional treatment process of cadmium and LiCl-KCl eutectic salt.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.293-294
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• 2005

The safety analyses and evaluation works on the process and facility for ACP demonstration have been performed. The several safety factors, such as the risk, environmental, radiation, structural, criticality, were analyzed. The analysis results confirmed the reliability of the safety on the ACP process and facility during normal and accident conditions.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.297-298
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• 2005

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.675-682
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• 2013

The sequential separation process, composed of an oxygen sparging process for separating lanthanides and a zone freezing process for separating Group I and II fission products, was evaluated and tested with a surrogate eutectic waste salt generated from pyroprocessing of used metal nuclear fuel. During the oxygen sparging process, the used lanthanide chlorides (Y, Ce, Pr and Nd) were converted into their sat-insoluble precipitates, over 99.5% at $800^{\circ}C$; however, Group I (Cs) and II (Sr) chlorides were not converted but remained within the eutectic salt bed. In the next process, zone freezing, both precipitation of lanthanide precipitates and concentration of Group I/II elements were preformed. The separation efficiency of Cs and Sr increased with a decrease in the crucible moving speed, and there was little effect of crucible moving speed on the separation efficiency of Cs and Sr in the range of a 3.7 - 4.8 mm/hr. When assuming a 60% eutectic salt reuse rate, over 90% separation efficiency of Cs and Sr is possible, but when increasing the eutectic salt reuse rate to 80%, a separation efficiency of about 82 - 86 % for Cs and Sr was estimated.

• 제어로봇시스템학회:학술대회논문집
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• 2004.08a
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• pp.840-845
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• 2004

The remote operation of the Advanced Spent Fuel Conditioning Process (ACP) is analyzed by using the 3D graphic simulation tools. Since the spent nuclear fuel, which is a high radioactive material, is processed in the ACP, the ACP equipment is operated in intense radiation fields as well as in a high temperature. Thus, the equipment is operated in a remote manner and should be designed with consideration for the remote handling and maintenance. Also suitable remote handling technology needs to be developed along with the design of the process concepts. For this we developed a graphic simulator, which provides the capability of verifying the remote operability of the ACP without the fabrication of the process equipment. In other words, by applying virtual reality to the remote maintenance operation, a remote operation task can be simulated in the graphic simulator, not in the real environment. The graphic simulator will substantially reduce the cost of the development of the remote handling and maintenance procedure as well as the process equipment, while at the same time developing a remote maintenance concept that is more reliable, easier to implement, and easier to understand.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.191-204
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• 2004

The research and development of effective management technologies of the spent fuels discharged from power reactors are an important and essential task of KAERI. In resent several years KAERI has focused on a project named "development and demonstration of the Advanced spent fuel Conditioning Process (ACP) in a laboratory scale." The Facility for ACP demonstration consists of two Hot Cells and auxiliary facilities. It is now in the final design stage and will be constructed in 2004. After construction of the facility the ACP equipments will be installed in Hot Cells. The ACP will be demonstrated by some simulated spent fuels first and then by spent fuels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.77-84
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• 2010

Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-$Li_2O$ molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.