Go to the main menu
Skip to content
Go to bottom
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 11, Issue 4 - Dec 2013
Volume 11, Issue 3 - Sep 2013
Volume 11, Issue 2 - Jun 2013
Volume 11, Issue 1 - Mar 2013
Selecting the target year
Sorptive Removal of Radionuclides (Cobalt, Strontium and Cesium) using AMP/IO-PAN Composites
Park, Younjin ; Kim, Chorong ; Shin, Won Sik ; Choi, Sang-June ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 259~269
DOI : 10.7733/jnfcwt-k.2013.11.4.259
Applicability of ammonium molybdophosphate/iron oxides-polyacrylonitrile (AMP/IO-PAN) composites on the removal of radionuclides in the radioactive wastewater generated from nuclear power plants was investigated. The composites were characterized using the following analytical techniques: X-ray diffraction (XRD), Fourior transform-infrared (FT-IR) spectroscopy, scanning electron microscopy (SEM), particle size analyzer (PSA), nitrogen adsorption-desorption and magnetic property measurement system (MPMS). 10wt% of AMP/IO-PAN composite has a saturation magnetization of 2.038 emu/g. Single-solute sorptions of Co, Sr and Cs onto 10wt% of AMP/IO-PAN composite were investigated. The maximum sorption capacities (
) predicted by the Langmuir model on 10wt% of AMP/IO-PAN composite were 0.097, 0.086 and 0.66 mmol/g for Co, Sr and Cs, respectively. The maximum sorption capacities (
) of Cs predicted by Langmuir model on 0, 10, 20 and 30wt% of AMP/IO-PAN composites were 0.702, 0.655, 0.602 and 0.559 mmol/g, respectively. The maximum sorption capacities (
) of Cs decreased with increasing the iron oxide content in the AMP/IO-PAN composites.
Characterization of Cement Waste Form for Final Disposal of Decommissioned Concrete Waste
Lee, Yoon Ji ; Hwang, Doo Seong ; Lee, Ki Won ; Jeong, Gyeong Hwan ; Moon, Jei Kwon ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 271~280
DOI : 10.7733/jnfcwt-k.2013.11.4.271
Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. And concrete waste was generated about 800 drums of 200 L. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. Thus, this research has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form to meet the requirements specified in disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10wt% as the optimized mixing ratio. Also, the compressive strength of cement waste form was satisfied that including fine powder up to maximum 40wt% in concrete debris wastes about 75%. As a result of scale-up test, the mixture of concrete waste, water, and cement is 75:10:15wt% meet the satisfied compressive strength because the free water increased with and increased in particle size.
Travel Times of Radionuclides Released from Hypothetical Multiple Source Positions in the KURT Site
Ko, Nak-Youl ; Jeong, Jongtae ; Kim, Kyung Su ; Hwang, Youngtaek ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 281~291
DOI : 10.7733/jnfcwt-k.2013.11.4.281
A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.
Borehole Disposal Concept: A Proposed Option for Disposal of Spent Sealed Radioactive Sources in Tanzania
Salehe, Mikidadi ; Kim, Chang-Lak ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 293~301
DOI : 10.7733/jnfcwt-k.2013.11.4.293
Borehole Disposal Concept (BDC) was initiated by the South African Nuclear Energy Corporation (NECSA) with the view to improve the radioactive waste management practices in Africa. At a time when geological disposal of radioactive waste is being considered, the need to protect ground water from possible radioactive contamination and the investigation of radionuclides migration through soil and rocks of zone of aeration into ground water has becomes very imperative. This is why the Borehole Disposal Concept (BDC) is being suggested to address the problem. The concept involves the conditioning and emplacement of disused sealed radioactive sources in an engineered facility of a relatively narrow diameter borehole (260 mm). Tanzania is operating a Radioactive Waste Management Facility where a number of spent sealed radioactive sources with long and short half lives are stored. The activity of spent sealed radioactive sources range from (1E-6 to 8.8E+3 Ci). However, the long term disposal solution is still a problem. This study therefore proposing the country to adopt the BDC, since the repository requires limited land area and has a low probability of human intrusion due to the small footprint of the borehole.
Deep Borehole Disposal Concept of Spent Fuel for Implementation in Korea
Yun, SooHyun ; Kim, Chang-Lak ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 303~309
DOI : 10.7733/jnfcwt-k.2013.11.4.303
As an alternative of the spent fuel disposal in a geologic repository, a deep borehole disposal concept for disposal at the section of 3 - 5km deep in a borehole has been proposed in several countries. In this paper, the latest reports of Sandia National Laboratories on the borehole disposal researches are analyzed. For implementation of this disposal concept in Korea, a conceptual design of spent fuel disposal canister and a modified deep borehole concept are suggested along with a required disposal area.
Concrete Degradation Comparison of Computer Programs for Post-Closure Safety Assessment of Wolsong Low-and Intermediate-Level Radioactive Waste Disposal Facility
Jung, Kang-Il ; Bang, Je-Heon ; Park, Jin Beak ; Yoon, Jeong Hyoun ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 311~324
DOI : 10.7733/jnfcwt-k.2013.11.4.311
To ensure the reliability of computer programs used for the post-closure safety assessment in the Wolsong LILW Center, the results from MASCOT, SAFE-ROCK and GOLDSIM programs are compared with a problem for degradation. Advantages and disadvantages of each computer programs are individually analyzed. Effects on the individual dose are assessed with each computer programs. MASCOT and SAFE-ROCK showed similar results for
. However, GOLDSIM represented different results for
. It is analyzed further and compared with the fluxes in each barrier of the disposal system. Througout the benchmarking testing of the computer program, the limitation of computer program can be continuously found out for the mature post-closure safety of Korean radwaste disposal system.
Separation and Solidification of Rare Earth Nuclides from LiCl-KCl Based Eutectic Waste Salts using a series of Phosphorylation/Distillation/Solidification Processes
Eun, Hee-Chul ; Choi, Jung-Hoon ; Cho, In-Hak ; Park, Hwan-Seo ; Park, Geun-Il ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 325~332
DOI : 10.7733/jnfcwt-k.2013.11.4.325
Pyroporcessing of spent nuclear fuel generates a considerable amount of LiCl-KCl eutectic waste salt containing radioactive rare earth (RE) chlorides. In this study, a series of processes, which consist of a phosphorylation/distillation process and a solidification process, were performed to minimize volume of the LiCl-KCl eutectic waste salt and solidify a residual waste into a stable form at a relatively low temperature. Over 99wt% of RE chlorides in LiCl-KCl eutectic salt was converted and separated into
in the phosphorylation/distillation process using a mixture of
. The separated
was solidified into a homogeneous and fine-grained form at
using LIP(Lead Iron Phosphate) as a solidification agent. The final waste volume was reduced below about 10% through the series of the processes.
Review on Spent Nuclear Fuel Performance and Degradation Mechanisms under Long-term Dry Storage
Kim, Juseong ; Kook, Donghak ; Sim, Jeehyung ; Kim, Yongsoo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 11, issue 4, 2013, Pages 333~349
DOI : 10.7733/jnfcwt-k.2013.11.4.333
As the capacity of spent nuclear fuel storage pool at reactor sites becomes saturated in ten years, long term dry storage strategy has been recently discussed as an alternative option in Korea. In this study, we reviewed safety-criteria-related research results on spent nuclear fuel performance and integrity under long-term dry storage and proposed the direction and the scope of future domestic research and development. Creep and hydride effect in relation to the embrittlement are known to be the major degradation mechanisms of the spent fuels during the long term dry storage. However, recent research results showed that hydride reorientation and hydride embrittlement are one of the most critical factors to the spent fuel integrity. Accordingly safety criteria of US and Japan for the storage system are basically founded on those mechanisms. However, in Korea, not only in-pile but out-of-pile experimental data have not been generated to understand fuel cladding degradation and to determine the criteria to ensure the safety. In addition, the transient behavior of the spent fuel during transportation also needs to be thoroughly examined. Therefore, various experimental research and development will be required to establish our own safety criteria for future long-term dry storage of domestic spent fuels.