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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
Journal Basic Information
Journal DOI :
The Korean Radioactive Waste Society
Editor in Chief :
Volume & Issues
Volume 8, Issue 4 - Dec 2010
Volume 8, Issue 3 - Sep 2010
Volume 8, Issue 2 - Jun 2010
Volume 8, Issue 1 - Mar 2010
Selecting the target year
A Study on the Fabrication of Uranium-Cadmium Alloy and its Distillation Behavior
Kim, Ji-Yong ; Ahn, Do-Hee ; Kim, Kwang-Rag ; Paek, Seung-Woo ; Kim, Si-Hyung ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 261~267
The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is "electrowinning" which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to
within a temperature range of 773 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller.
Multilateral Approaches to the Back-end of the Nuclear Fuel Cycle: Challenges and Possibilities
Ryu, Ho-Jin ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 269~277
Various multilateral approaches to the nuclear fuel cycle have been proposed in order to suppress the expansion of sensitive fuel cycle technology. In order to prepare for the future multilaterallization of the nuclear fuel cycle, existing multilateral spent fuel management programs are analyzed. A trial multilateralization of a domestic R&D facility for the back end of the nuclear fuel cycle is proposed and its challenges, possibilities and implementation strategy are discussed.
Study on the Geological Structure around KURT Using a Deep Borehole Investigation
Park, Kyung-Woo ; Kim, Kyung-Su ; Koh, Yong-Kwon ; Choi, Jong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 279~291
To characterize geological features in study area for high-level radioactive waste disposal research, KAERI (Korea Atomic Energy Research Institute) has been performing the several geological investigations such as geophysical surveys and borehole drilling since 1997. Especially, the KURT (KAERI Underground Research Tunnel) constructed to understand the deep geological environments in 2006. Recently, the deep borehole of 500 m depths was drilled to confirm and validate the geological model at the left research module of the KURT. The objective of this research was to identify the geological structures around KURT using the data obtained from the deep borehole investigation. To achieve the purpose, several geological investigations such as geophysical and borehole fracture surveys were carried out simultaneously. As a result, 7 fracture zones were identified in deep borehole located in the KURT. As one of important parts of site characterization on KURT area, the results will be used to revise the geological model of the study area.
Assessment on the Monitoring System for KURT using Optical Fiber Sensor Cable
Kim, Kyung-Su ; Bae, Dae-Seok ; Koh, Yong-Kwon ; Kim, Jung-Yul ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 293~301
Optical fiber cable, as a sensor, was installed on the wall of KAERI(Korea Atomic Energy Research Institute) Underground Research Tunnel(KURT) in order to monitor the physical stability of the tunnel, which was constructed for technical development and demonstration of radioactive waste disposal. This monitoring system has two simultaneous measurements of temperature and strain over time using Brillouin backscatter. According to the results of the monitoring from Jan. 2008 to Nov. 2009, there is no significant displacement or movement at the tunnel wall However, the cumulative volume of total strain increased slightly as time passes with the comparison of the reference observation, which was measured in Jan. 2008. The change in cumulative volume of total strain indicates that the strain level had been affected by saturation and de-saturation phenomena due to groundwater fluctuation at several points at KURT. This system is based on the distributed sensing technique concept, not point sensing. By using this system, a displacement can be detected with the range from
every 1m interval in minimum. A temperature variation can be monitored at every 0.5m interval with the resolution of 0.01 in minimum. Based on the study, this monitoring system is potentially applicable to long term monitoring systems for radioactive waste disposal project as well as other structures and underground openings.
Recovery of Residual LiCl-KCl Eutectic Salts in Radioactive Rare Earth Precipitates
Eun, Hee-Chul ; Yang, Hee-Chul ; Kim, In-Tae ; Lee, Han-Soo ; Cho, Yung-Zun ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 303~309
For the pyrochemical process of spent nuclear fuels, recovery of LiCl-KCl eutectic salts is needed to reduce radioactive waste volume and to recycle resource materials. This paper is about recovery of residual LiCl-KCl eutectic salts in radioactive rare earth precipitates (rare earth oxychlorides or oxides) by using a vacuum distillation process. In the vacuum distillation test apparatus, the salts in the rare earth precipitates were vaporized and were separated effectively. The separated salts were deposited in three positions of the vacuum distillation test apparatus or were collected in the filter and it is difficult to recover them. To resolve the problem, a vacuum distillation and condensation system, which is subjected to the force of a temperature gradient at a reduced pressure, was developed. In a preliminary test of the vacuum distillation/condensation recovery system, it was confirmed that it was possible to condense the vaporized salts only in the salt collector and to recover the condensed salts from the salt collector easily.
A Study on Improvement of Test Method of Nuclear Power Plant ESF ACS by applying Regulatory Guide 1.52 (Rev.3)
Lee, Sook-Kyung ; Kim, Kwang-Sin ; Sohn, Soon-Hwan ; Song, Kyu-Min ; Lee, Kae-Woo ; Park, Jeong-Seo ; Cho, Byoung-Ho ; Yoo, Byeang-Jea ; Hong, Soon-Joon ; Kang, Sun-Haeng ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 311~318
U. S. NRC Regulation Guide 1.52 regulating ESF ACS in nuclear power plants has been revised to revision 3. To apply reduction of operability test time, allowance of alternative challenge agents for in-place leak test of HEPA filters, and upgrade of Methyl Iodide penetration acceptance criterion in activated carbon performance test suggested in Reg. Guide 1.52(Rev.3) on Yonggwang units 5 and 6 ESF ACSes, technical feasibility study was carried out with on-site experiments as well as experiments with a lab-scale model. It was confirmed that the moisture in the system returned to the level before the test in 1 or 4 days even though the moisture was removed during the operability test lasting more than 10 hours. Therefore, it is appropriate to perform monthly operability test in 15 minutes just long enough to check the operability of equipment. To change challenge material for in-place HEPA filter leak test, size of aerosol, production rate, and leak detection capability were compared for DOP and PAO. It was concluded that PAO can be substituted for DOP in nuclear power plants. The upgrade of Methyl Iodide penetration acceptance criterion from 0.175 % to 0.5 % in active carbon filter bed deeper than 4 inches was to conform to the change of activated carbon performance test method to ASTM D3803(1989). It was confirmed that Methyl Iodide penetration acceptance criterion of 0.5 % under
, relative humidity 95 % condition was conservatively good enough for testing performance of active carbon insitu. The licence change of Yonggwang units 5 and 6 has been completed based on this study.
A Study on Construction and Application of Nuclear Grade ESF ACS Simulator
Lee, Sook-Kyung ; Kim, Kwang-Sin ; Sohn, Soon-Hwan ; Song, Kyu-Min ; Lee, Kei-Woo ; Park, Jeong-Seo ; Hong, Soon-Joon ; Kang, Sun-Haeng ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 319~327
A nuclear plant ESF ACS simulator was designed, built, and verified to perform experiment related to ESF ACS of nuclear power plants. The dimension of 3D CAD model was based on drawings of the main control room(MCR) of Yonggwang units 5 and 6. The CFD analysis was performed based on the measurement of the actual flow rate of ESF ACS. The air flowing in ACS was assumed to have
and uniform flow. The flow rate across the HEPA filter was estimated to be 1.83 m/s based on the MCR ACS flow rate of 12,986 CFM and HEPA filter area of 9 filters having effective area of
each. When MCR ACS was modeled, air flow blocking filter frames were considered for better simulation of the real ACS. In CFD analysis, the air flow rate in the lower part of the active carbon adsorber was simulated separately at higher than 7 m/s to reflect the measured value of 8 m/s. Through the CFD analyses of the ACSes of fuel building emergency ventilation system, emergency core cooling system equipment room ventilation cleanup system, it was confirmed that all three EFS ACSes can be simulated by controlling the flow rate of the simulator. After the CFD analysis, the simulator was built in nuclear grade and its reliability was verified through air flow distribution tests before it was used in main tests. The verification result showed that distribution of the internal flow was uniform except near the filter frames when medium filter was installed. The simulator was used in the tests to confirm the revised contents in Reg. Guide 1.52 (Rev. 3).
Soil-to-Rice Seeds Transfer Factors of Radioiodine and Technetium for Paddy Fields around the Radioactive-Waste Disposal Site in Gyeongju
Choi, Yong-Ho ; Lim, Kwang-Muk ; Jun, In ; Park, Doo-Won ; Keum, Dong-Kwon ; Han, Moon-Hee ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 329~337
Radiotracer experiments were performed over two years using pot cultures in a greenhouse to investigate soil-torice seeds transfer factors of radioiodine and technetium for paddy fields around the radioactive-waste disposal site in Gyeongju. Before transplanting rice seedlings, the top about 20 cm soils were thoroughly mixed with
(2008), and the pots were irrigated to simulate flooded rice fields. Transfer factors were determined as the ratios of the radionuclide concentrations in dry rice seeds (brown rice) to those in dry soils. Transfer factors of radioiodine and technetium were in the ranges of
(three soils) and
(four soils), respectively, for different soils. It seems that the differences in the clay content among soils played a more important role for such variations than those in the organic matter content and pH. As the representative values of radioiodine and technetium transfer factors for rice seeds,
, respectively, were proposed. In order to obtain more highly representative values in the future, investigations for the sites of interest need to be carried out continuously.
Assessment of Potential Impacts of the Proposals for Multilateralization of Nuclear Fuel Cycle
Moon, Joo-Hyun ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 339~346
Recently, there have been grave concerns that the anticipated increase in the use of nuclear energy worldwide could result in dissemination of sensitive nuclear technologies. To meet the increase in nuclear energy demand and strengthen the non-proliferation regime simultaneously, the various proposals for 'multilateralization of nuclear fuel cycle' have been widely suggested. Those proposals are expected to have serious impacts on our country, if they has come true. In this paper, therefore, the 12 existing proposals were reviewed and assessed for their potential impacts on our country, in order to help prepare the appropriate measures responding to the international attempt of 'multilateralization of nuclear fuel cycle'.
Source Term Characterization for Structural Components in
KOFA Spent Fuel Assembly
Cho, Dong-Keun ; Kook, Dong-Hak ; Choi, Heui-Joo ; Choi, Jong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 4, 2010, Pages 347~353
Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be
Bequerels, 236 Watts,
-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20~45 % and 30~45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.