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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 6, Issue 4 - Dec 2008
Volume 6, Issue 3 - Sep 2008
Volume 6, Issue 2 - Jun 2008
Volume 6, Issue 1 - Mar 2008
Selecting the target year
Reductive stripping of Np using a n-butyraldehyde from a loaded TBP phase containing Np
Lee, Eil-Hee ; Lim, Jae-Kwan ; Chung, Dong-Yong ; Yang, Han-Beom ; Kim, Kwang-Wook ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 163~170
The reductive stripping of Np using a n-butyraldehyde (NBA) from loaded organic solution containing Np, which was oxidative-extracted in a system of a 30 % TBP/NDD-2M
and O/A=2 containing 0.005 M
as an oxidant of Np, was studied. The stripping yields of Np was increased with an increasing the NBA concentration, with a decreasing the nitric acid concentration of stripping solution and with a decreasing the reaction temperature. The apparent reductive stripping rate equation was shown by the following equation :
= 1,524 exp(-2,906/T)
. At 1.04 M NBA and 2 M
, the stripping yield of Np and U was 70.1 %, and 7.1 %, respectively, and the separation factor of U over Np (
) was about 30.4. Therefore, it was found that U and Np co-extracted in a system of TBP-
could be effectively mutual-separated by the NBA.
Evaluation of Dark Spots Formated on the High Temperature Metal Filter Elements
Park, Seung-Chul ; Hwang, Tae-Won ; Moon, Chan-Kook ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 171~178
Metal filter elements were newly introduced to the high temperature filter(HTF) system in the low- and intermediate-level radioactive waste vitrification plant. In order to evaluate the performance of various metal materials as filter media, elements made of AISI 316L, AISI 904L, and Inconel 600 were included to the test set of filter elements. At the visual inspection to the elements performed after completion of each test, a few dark spots were observed on the surface of some elements. Especially they were found much more at the AISI 316L elements than others. To check the dark spots are the corrosion phenomena or not, two kinds of analyses were performed to the tested filter elements. Firstly, the surfaces or the cross sections of filter specimens cut out from both normal area and dark spot area of elements were analyzed by SEM/EDS. The results showed that the dark spots were not evidences of corrosion but the deposition of sodium, sulfur and silica compounds volatilized from waste or molten glass. Secondly, the ring tensile strength were analyzed for the ring-shape filter specimens cut out from each kind of element. The result obtained from the strength tested showed no evidence of corrosion as well. Conclusionally, depending on the two kinds of analysis, no evidences of corrosion were found at the tested metal filter elements. But the dark spots formed on the surface could reduce the effective filtering area and increase the overall pressure drop of HTF system. Thus, continuous heating inside filter housing up to dew point will be required normally. And a few long-period test should be followed for the exact evaluation of corrosion of the metal filter elements.
Canadian Public and Stakeholder Engagement Approach to a Spent Nuclear Fuel Management
Hwang, Yong-Soo ; Kim, Youn-Ok ; Whang, Joo-Ho ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 179~187
After Canada has struggled with a radioactive waste problem over for 20 years, the Canadian government finally found out that its approach by far has been lack of social acceptance, and needed a program such as public and stakeholder engagement (PSE) which involves the public in decision-making process. Therefore, the government made a special law, called Nuclear Fuel Waste Act (NFWA), to search for an appropriate nuclear waste management approach. NFWA laid out three possible approaches which were already prepared in advance by a nuclear expert group, and required Nuclear Waste Management Organization (NWMO) to be established to report a recommendation as to which of the proposed approaches should be adopted. However, NFWA allowed NWMO to consider additional management approach if the other three were not acceptable enough. Thus, NWMO studied and created a fourth management approach after it had undertaken an comparison of the benefits, risks and costs of each management approach: Adaptive Phased Management. This approach was intended to enable the implementers to accept any technological advancement or changes even in the middle of the implementation of the plan. The Canadian PSE case well shows that technological R&D are deeply connected with social acceptance. Even though the developments and technological advancement are carried out by the scientists and experts, but it is important to collect the public opinion by involving them to the decision-making process in order to achieve objective validity on the R&D programs. Moreover, in an effort to ensure the principles such as fairness, public health and safety, security, and adoptability, NWMO tried to make those abstract ideas more specific and help the public understand the meaning of each concept more in detail. Also, they utilized a variety of communication methods from face-to-face meeting to e-dialogue to encourage people to participate in the program as much as possible. Given the fact that Korea has been also having a hard time in dealing with spent nuclear fuel management, all of these efforts that Canada has made with a PSE program would give good lessons and implications to the Korean case. In conclusion, as a deliberative participation program, PSE could be a possible breakthrough approach for the Korean spent nuclear fuel management.
Evaluation of X-ray System for Nondestructive Testing on Radioactive Waste Drums
Park, Jong-Kil ; Maeng, Seong-Jun ; Lee, Yeon-Ee ; Hwang, Tae-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 189~203
The physical and chemical properties of radioactive waste drums, which have been temporarily stored on site, should be characterized before their shipment to a disposal facility in order to prove that the properties meet the acceptance guideline. The investigation of NDT(Nondestructive Test) method was figured out that the contents in drum, the quantitative analysis of free standing water and void fraction can be examined with X-ray NDT techniques. This paper describes the characteristics of X-ray NDT such as its principles, the considerations for selection of X-ray system, etc. And then, the waste drum characteristics such as drum type and dimension, contents in drum, etc. were examined, which are necessary to estimate the optimal X-ray energy for NDT of a drum. The estimation results were that:
the proper X-ray energy is under 3 MeV to test the drums of 320
both X-ray systems of 450 keV and/or 3 MeV might be needed considering the economical efficiency and the realization. The number of drums that can be tested with 450 keV and 3 MeV X-ray system was figured out as 42,327 and 18,105 drums (based on storage of 2006. 12), respectively. Four testing scenarios were derived considering equipment procurement method, outsourcing or not, etc. The economical and feasibility assessment for the scenarios was resulted in that an optimal scenario is dependent on the acceptance guide line, the waste generator's policy on the waste treatment and the delivery to a disposal facility, etc. For example, it might be desirable that a waste generator purchases two 450 keV mobile system to examine the drums containing low density waste, and that outsourcing examination for the high density drums, if all NDT items such as quantitative analysis for 'free standing water' and 'void fraction', and confirmation of contents in drum have to be characterized. However, one 450 keV mobile system seems to be required to test only the contents in 13,000 drums per year.
Development and Application of SITES
Park, Joo-Wan ; Yoon, Jeong-Hyoun ; Kim, Chank-Lak ; Cho, Sung-Il ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 205~215
SITES(Site Information and Total Environmental Data Management System) has been developed for the purpose of systematically managing site characteristics and environmental data produced during the pre-operational, operational, and post-closure phases of a radioactive waste disposal facility. SITES is an integration system, which consists of 4 modules, to be available for maintenance of site characteristics data, for safety assessment, and for site/environment monitoring; site environmental data management module(SECURE), integrated safety assessment module(SAINT), site/environment monitoring module(SUDAL) and geological information module for geological data management(SITES-GIS). Each module has its database with the functions of browsing, storing, and reporting data and information. Data from SECURE and SUDAL are interconnected to be utilized as inputs to SAINT. SAINT has the functions that multi-user can access simultaneously via client-server system, and the safety assessment results can be managed with its embedded Quality Assurance feature. Comparison between assessment results and environmental monitoring data can be made and visualized in SUDAL and SITES-GIS. Also, SUDAL is designed that the periodic monitoring data and information could be opened to the public via internet homepage. SITES has applied to the Wolsong low- and intermediate-level radioactive waste disposal center in Korea, and is expected to enhance the function of site/environment monitoring in other nuclear-related facilities and also in industrial facilities handling hazardous materials.
Analysis of the Thermal and Structural Stability for the CANDU Spent Fuel Disposal Canister
Lee, Jong-Youl ; Cho, Dong-Geun ; Kim, Seong-Gi ; Choi, Heui-Joo ; Lee, Yang ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 217~224
In deep geological disposal system, the integrity of a disposal canister having spent fuels is very important factor to assure the safety of the repository system. This disposal canister is one element of the engineered barriers to isolate and to delay the radioactivity release from human beings and the environment for a long time so that the toxicity does not affect the environment. The main requirement in designing the deep geological disposal system is to keep the buffer temperature below 100
by the decay heat from the spent fuels in the canister in order to maintain the integrity of the buffer material. Also, the disposal canister can endure the hydraulic pressure in the depth of 500 m and the swelling pressure of the bentonite as a buffer. In this study, new concept of the disposal canister for the CANDU spent fuels which were considered to be disposed without any treatment was developed and the thermal stability and the structural integrity of the canister were analysed. The result of the thermal analysis showed that the temperature of the buffer was 88.9
when 37 years have passed after emplacement of the canister and the spacings of the disposal tunnel and the deposition holes were 40 m and 3 m, respectively. In the case of structural analysis, the result showed that the safety factors of the normal and the extreme environment were 2.9 and 1.33, respectively. So, these results reveal that the canister meets the thermal and the structural requirements in the deep geological disposal system.
Reference Spent Nuclear Fuel for Pyroprocessing Facility Design
Cho, Dong-Keun ; Yoon, Seok-Kyun ; Choi, Heui-Joo ; Choi, Jong-Won ; Ko, Won-Il ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 225~232
An estimation has been made for inventories and characteristics of spent nuclear fuel(SNF) to be generated from existing and planned nuclear power plants based on the 3rd Basic Plan for Electric Power Demand and Supply. The characteristics under consideration in this study are dimensions, a fuel rod array, a weight,
enrichment, and the discharge burnup in terms of fuel assembly. These are essentially needed for designing a pyroprocessing facility. It is appeared that the anticipated quantity by the end of 2077 is about 23,000 tU for PWR spent nuclear fuel. It is revealed that the proportion of SNF with the initial
enrichment below 4.5 weight percent(wt.%) is approximately 95 % in total. For SNF with 16
16 fuel rod array the proportion is expected approximately 74% in total. It appears that the average burnup of SNF will be 55 GWd/tU after the medium and/or latter part of 2010s while the average burnup is 45 GWd/tU at present. Finally, a requirement in terms of reference SNF for designing the pyroprocessing facility has been derived from the above-mentioned results. The anticipated SNF seems to be 16
16 Korean Standard Fuel Assembly with a cross section of 21.4 cm
21.4 cm, a length of 453 cm, a mass of 672 kg, the initial
enrichment of 4.5 wt.%, and the discharge burnup of 55 GWd/tU.
A Conceptual Design Study for a Spent Fuel Pyroprocessing Facility of a Demonstration Scale
Yoo, Jae-Hyung ; Hong, Kwon-Pyo ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 3, 2008, Pages 233~244
A conceptual design study for a pyroprocesing facility, has been carried out in this study, which is available for the recovery of uranium and transuranic elemental group(TRU), that is, reusable as a nuclear fuel especially in a next generation-type reactor. The scale of this facility has been chosen as 20 kg HM/batch, comparatively small engineering size in order to collect scale-up data for the design of a commercial facility as well as to get operational experience. The spent fuel to be handled in this process is as follows : 3.5 % enriched uranium fuel, 35,000MWd/tU and 5-year cooled. The major items considered in the conceptual study are a building lay-out including various hot cells, safety management of the process operation in conjunction with material balance control, radiation safety, inert atmosphere control in shielded hot cells, and criticality control of uranium and TRU products.