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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 4, Issue 4 - Dec 2006
Volume 4, Issue 3 - Sep 2006
Volume 4, Issue 2 - Jun 2006
Volume 4, Issue 1 - Mar 2006
Selecting the target year
A Study on the Wigner Energy Release Characteristics of Irradiated Graphite of KRR-2
Jeong Gyeong-Hwan ; Yun Sei-Hun ; Lee Dong-Gyu ; Jung Chong-Hun ; Lee Keun-Woo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 209~216
Characteristics of heat release process, while the Wigner energy was drawn off the graphite during DSC(Differential Scanning Calorimenter) measurement as an example of annealing process which is one of release methods of Wigner energy that is contained in the irradiated graphite, was studied. Linear temperature rise method in DSC operation was selected to estimate the total Wigner energy content and the heat release rate of each graphite samples, which were located in several positions in the thermal column in KRR-2 research reactor. As an annealing process in DSC operation Wigner energy of the irradiated graphite samples were totally released by heat supplying to the graphite from room temperature to
, in DSC. Characteristics of Wigner energy release from the graphite sample was well correlated with the various activation energy model of the kinetic equation.
The Development and Performance Evaluation of a Cyclone to Remove Hot Particulate from a Contaminated Hot Cell
Kim Gye-Nam ; Won Hui-Jun ; Choi Wang-Kyu ; Jung Chong-Hun ; Oh Won-Zin ; Park Jin-Ho ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 217~226
The structural and contamination characteristics of hot cells at KAERI were investigated. The SEM results showed that the size of the hot particulate on the inner surface of the hot cell ranged from 0.2 to
. It was found that an inlet flow rate of 15 m/sec was suitable for this developed cyclone with a 49 mm optimum vortex finder length. The results showed that the collection efficiency was about 85% for
particles. The collection efficiency didn't show a sharp increase when the inlet flow rate was faster than 15m/sec. When the temperature of the inlet flow gas was increased, the collection efficiency of the cyclone was slightly decreased. The larger the vortex finder length was, the higher the pressure drop in the cyclone was. The cut size diameter decreased with an increment of the Reynolds number. It was established that the flow in the cyclone was a turbulent flow on the basis of the Reynolds number and this turbulent flow caused a pressure drop in the cyclone.
decreased with increasing values of the Reynolds number and it gradually approached a constant value at a higher value of the Reynolds number Namely,
approached approximately 0.045 between 6000 and 8000 of the Reynolds number.
Evaluation of Chemical Durability of Vitrified Forms for Simulated Radioactive Waste Using Product Consistency Test(PCT) and Vapor Hydration Test(VHT)
Kim Cheon-Woo ; Kim Ji-Yean ; Maeng Sung-Jun ; Park Jong-Kil ; Hwang Tae-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 227~234
Two candidate glasses, AG8W1 and DG-2, have been developed for vitrifying the mixture of low activity resin, zeolite and Dry Active Waste(DAW), and DAW solely, respectively. In order to evaluate the chemical durability of the glasses, two different leaching tests, Product Consistency Test(PCT) and Vapor Hydration Test(VHT), have been performed. As the results of PCT performed from 7 to 120 days, the leach rates of B, Na, Si and Li in the glasses were much lower than those of the benchmark glass(SRL-EA). As the result of VHT peformed for 7 days, the leach rates were 2 and
for AG8W1 and DG-2, respectively, The results of VHT met the regulatory guideline( $<50g/m^2/day$) for the low activity glasses of Hanford in the USA. Consequently, two candidate glasses to be used at a commercial operation in the future showed that their chemical durability is satisfactory according to the results of two leaching tests.
An Experimental Study on the Sorption of Uranium(VI) onto a Bentonite Colloid
Baik Min-Hoon ; Cho Won-Jin ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 235~243
In this study, an experimental study on the sorption properties of uranium(VI) onto a bentonite colloid generated from Gyeongju bentonite which is a potential buffer material in a high-level radioactive waste repository was performed as a function of the pH and the ionic strength. The bentonite colloid prepared by separating a colloidal fraction was mainly composed of montmorillonite. The concentration and the size fraction of the prepared bentonite colloid measured using a gravitational filtration method was about 5100 ppm and 200-450 nm in diameter, respectively. The amount of uranium removed by the sorption reaction bottle walls, by precipitation, and by ultrafiltration was analyzed by carrying out some blank tests. The removed amount of uranium was found not to be significant except the case of ultrafiltration at 0.001 M
. The ultrafiltration was significant in the lower ionic strength of 0.001 M
due to the cationic sorption onto the ultrafilter by a surface charge reversion. The distribution coefficient
(or pseudo-colloid formation constant) of uranium(VI) for the bentonite colloid was about
depending upon pH and ionic strength of
was highest in the neutral pH around 6.5. It is noted that the sorption of uranium(VI) onto the bentonite colloid is closely related with aqueous species of uranium depending upon geochemical parameters such as pH, ionic strength, and carbonate concentration. As a consequence, the bentonite colloids generated from a bentonite buffer can mobilize the uranium(VI) as a colloidal form through geological media due to their high sorption capacity.
The Test for Verifying a Tip-Over Analysis of a Dry Storage Cask
Kim Dong-Hak ; Seo Ki-Seog ; Lee Ju-Chan ; Cho Chun-Hyung ; Jang Hyun-Kee ; Choi Byung-Il ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 245~253
A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry. concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the non-detective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tip-over test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.
Manufacture of the vol-oxidizer with a capacity of 20 kg HM/batch in
pellets using a design model
Kim Young-Hwan ; Yoon Ji-Sup ; Jung Jae-Hoo ; Hong Dong-Hee ; Uhm Jae-Beop ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 255~263
Vol-oxidizer is a device to convert
powder and to feed a homogeneous powder into a Metal Conversion Reactor in the ACP(Advanced Spent Fuel Conditioning Process). In this paper, we propose a design model of the vol-oxidizer, develop the new vol-oxidizer with a capacity of 20 kg HM/batch in
pellets, and conduct a verification for the device. Design considerations include the internal structure, the capacity, the heating position of the device, and the size. The dimensions of the new vol-oxidizer are decided by the design model. We determine a permeability test of the
measuring the temperature distribution, and the volume of
. We manufactured the new vol-oxidizer for a 20 kg HM/batch in
pellets, and then analyzed the characteristics of the
powder for the verification. The experimental results show that the permeability of the
throughout mesh enhance more than old vol-oxidizer, the oxidation time takes only 8 hours when compared with the 13 hours of the old device, and the average distribution of particle size is
. The capacities of new vol-oxidizer for a 20 kg HM/batch in
pellets were agree well with the predictions of design model.
Development of Evaluation Modules for Evaluating Decommissioning Scenarious Using Digital Mock-Up System
Kim Sung-Kyun ; Park Hee-Sung ; Lee Kune-Woo ; Jung Chong-Hun ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 265~273
In the decommissioning and decontamination(D&D) planning stage, it is important that the scenarios are evaluated from an engineering point of views because the decommissioning work has to be executed economically and safely by following the best scenarios. Therefore, we need to develope several modules to evaluate the decommissioning scenarios. In this paper, the digital mock-up system is constructed in the virtual space to simulate the whole decommissioning process. The schedule evaluating equation and cost evaluation equation are derived to calculate the working time and the expected cost. And in order to easily identify the radiation level about the activated objects, the radiation visualization module is developed. Finally, on the basis of the obtained results from the Digital Mock-up and other important factors, the evaluating method of the scenarios that can indicate the best scenario is described.
Comparison of the Correction Methods for Gamma Ray Attenuation in the Radioactive Waste Drum Assay
Ji Young-Yong ; Ryu Young-Gerl ; Kwak Kyoung-Kil ; Kang Duck-Won ; Kim Ki-Hong ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 275~284
In the measurement of gamma rays emitted from the nuclide in the radioactive waste drum, to analyze the nuclide concentration accurately, it is necessary to use the proper calibration standards and to correct for the attenuation of the gamma rays. Two drums having a different density were used to analyze the nuclide concentration inside the drum in this study. After carrying out the system calibration, we measured the gamma rays emitted from the standard source inside the model drum with changing the distance between the drum and the detector. The measured values were corrected with the three kinds of gamma attenuation correction methode, as a results, the error was less than 10 % in the low density drum and less than 25 % in the high density drum. The measured activity in the short distance was more accruable than in the long distance. The transmission correction for the mass attenuation showed good results(very Low error) compared to the mean density and the differential peak correction method.
National Policy and Status on Management of Spent Nuclear Fuel
Park Won-Jae ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 285~299
At the end of 2005, 443 nuclear reactors were operating in 32 countries worldwide. They had provided about 3,000 TWh, which was just over 16 percent of global electricity supply. With the generating capacity of 368 GWe in 2004, the spent fuel generation rate worldwide, now becomes at about 11,000 tHM/y. Projections indicate that cumulative amounts to be generated by the year 2020, the time when most of the existing NPP will be closed to the end of their licensed lifetime, may be close to 445,000 tHM. In this regard, spent fuel management is a common issue in all countries with nuclear reactors. Whatever their national policy and/or strategy is selected for the backend of the nuclear fuel cycle, the management of spent fuel will contribute an impending and imminent issues to be resolved in the foreseeable future. The 2nd Review Meeting of the Contracting Parties to the Joint Convention was held in Vienna from 15 to 24 May 2006. The meeting gave an opportunity to exchange information on the national policy and strategy of spent fuel management of the Contracting Parties, to discuss their situations, prospects and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should be taken. In this paper, an overview of national and global trends of spent fuel management is discussed. In addition, some directions are identified and recent activities of each Member States in the subject area are summarized.
Structural Analysis of the Canister for PWR Spent Fuels under the Korean Reference Disposal Conditions
Choi Heui-Joo ; Lee Yang ; Choi Jong-Won ; Kwon Young-Joo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 3, 2006, Pages 301~309
KDC-1 canister for PWR spent fuels which will be used for the Korean Reference Disposal System was developed. The structural analysis of the canister was carried out as a part of the safety analysis. Two conditions, disposal condition and handling condition, were considered for the structural analysis. Three kinds of load cases, normal, abnormal and rock movement, were considered for the disposal condition. The results of the calculation showed that the safety factors from the structural analysis were greater than the design requirements. Two accident scenarios, gripper failure accident and canister drop accident, were analyzed for the handling condition. According to the gripper failure scenario analysis, the handling machine with grippers could be used even in the cases that one or two grippers failed. The maximum von Mises stress from the canister drop accident scenario was 0.762 MPa, which was negligible compared with the yield stress of nodular cast iron. The proposed KDC-1 canister for PWR spent fuels proves to be safe under the repository condition that is based upon the Korean reference disposal system according to the structural analysis for disposal condition and handling condition.