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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
Biosphere Modeling for Dose Assessment of HLW Repository: Development of ACBIO
Lee, Youn-Myoung ; Hwang, Yong-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 73~100
For the purpose of evaluating dose rate to individual due to long-term release of nuclides from the HLW repository, a biosphere assessment model and the implemented code, ACBIO, based on BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To show its practicability and usability as well as to see the sensitivity of compartment scheme or parametric variation to concentration and activity in compartments as well as annual flux between compartments at their peak values, some calculations are made and investigated: For each case when changing the structure of compartments and GBIs as well as varying selected input Kd values, all of which seem very important among others, dose rate per nuclide release rate is separately calculated and analyzed. From the maximum dose rates (Bq/y), flux-to-dose conversion factors (Sv/Bq) for each nuclide were derived, which are to be used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rate (Sv/y) for individual in critical group. It has been also observed that compartment scheme, identification of possible exposure group and GBIs could be all highly sensitive to the final consequences in biosphere modeling.
Thermal Conductivity of Compacted Bentonite and Bentonite-Sand Mixture
Cho, Won-Jin ; Lee, Jae-Owan ; Kwon, Sang-Ki ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 101~109
For the Kyungju bentonite which is considered as a candidate material for the buffer and backfill in the high-level waste repository, the thermal conductivities of compacted bentonite and a bentonite-sand mixture were measured. The thermal conductivities of the compacted bentonites with a dry density of 1.2 to
and the bentonite-sand mixture with a dry density of 1.6 and
were measured within the gravimetric water content range of 10wt% to 20wt% and the sand fraction range of 10 to 30wt%. The thermal conductivity of compacted bentonite and a bentonite-sand mixture increases with increasing dry density and sand weight fraction in the case of constant water weight fraction, and increases with increasing water weight fraction and sand weight fraction in the case of constant dry density. The empirical correlations to describe the thermal conductivity of compacted bentonite and a bentonite-sand mixture as a function of water fraction at each dry density were suggested. These correlations can predict the thermal conductivities of bentonite and a bentonite-sand mixture with a difference below 10%.
Phoswich Detector for Simultaneous Measuring Alpha/beta Particles
Kim, Gye-Hong ; Park, Chan-Hee ; Lee, Kune-Woo ; Jung, Chong-Hun ; Seo, Bum-Kyoung ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 111~117
The new type phoswich detector consisting of the ZnS(Ag) and plastic scintillator for alpha/beta-ray simultaneous counting was developed for monitoring radiological contamination inside pipes. The detection performance was estimated using the PSD (pulse shape discrimination) method as a function of distance between the scintillator and radioactive source. The attenuation of particles traveling through a thin film for preventing the detector from being contaminated was experimentally estimated. It is concluded from our investigation that the phoswich detector developed can provide a sufficient alpha/beta-ray discrimination. The application of a thin film for preventing the detector from being contaminated was proven to be feasible.
Thermal Analysis of a Retrievable CANDU Spent Fuel Disposal Tunnel
Cha, Jeong-Hun ; Lee, Jong-Youl ; Choi, Heui-Joo ; Cho, Dong-Keun ; Kim, Sang-Nyung ; Youn, Bum-Soo ; Ji, Joon-Suk ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 119~128
Thermal assessment of a new CANDU spent fuel disposal system, which improves the retrievability of the spent fuel and enhances the densification factor compared with the Korean Reference disposal System, is carried out in this study. The canisters for CANDU spent fuels are stored for long term and cooled by natural convection in the proposed disposal system for the retrievability. The steady state thermal analyses for proposed CANDU disposal system are carried out with the ANSYS 10.0 CFX code. The thermal analyses are performed through two steps. At the first step, the sensitivity of the disposal tunnel spacing is analysed. The differences of maximum temperatures by several tunnel spacings are calculated at three points in the disposal tunnel. The result shows that the differences of the temperature at the three points are almost negligible because 99% of the decay heat is removed by natural convection. At the second procedure, 60m tunnel spacing with a ventilation system instead of natural convection is considered. The result is applied to the calculation of the canister surface temperature in disposal tunnel as boundary conditions. Consequently, the average and the maximum surface temperature of disposal canisters are
, respectively. The inner maximum temperature of a basket in the disposal canister is calculated as
. The maximum temperature of the basket meets the thermal requirement for the CANDU spent fuel cladding.
Sorption of Eu(III) and Th(IV) on Bentonite Colloids Considering Their Precipitation and Colloid Formation
Baik, Min-Hoon ; Lee, Jae-Kwang ; Lee, Seung-Yeop ; Kim, Seung-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 129~139
In this study, a sorption experiment of multivalent nuclides such as Eu(III) and Th(IV) relatively stable for redox reactions was carried out for bentonite colloids which had been prepared from the domestic Gyeongju bentonite. The amounts of the nuclides lost by an attachment to bottle walls, by a precipitation, and by a colloid formation were estimated by performing blank tests for the sorption experiments. Sorption coefficients,
, reflecting the mass losses were obtained and investigated for the sorption of Eu(III) and Th(IV) onto the bentonite colloids. The net sorption coefficients
considering all the three mass losses were measured as about
for Eu(III) and Th(IV), respectively, depending on pH. In particular, a precipitation occurred mainly at a pH greater than 5 for Eu(III) and a precipitation and colloid formation significantly occurred at a pH greater than 3 for Th(IV). The precipitation and colloid formation of the multivalent nuclides of Eu(III) and Th(IV) therefore should be considered when
are rightly obtained over the pH range where their precipitation and colloid formation become significant at a given concentration.
Uncertainty Cases in Economic Evaluation of Back-End Nuclear Fuel Cycle
Kim, Hyung-Joon ; Cho, Chun-Hyung ; Lee, Kyung-Ku ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 141~145
Due to the uncertainties resulting from cost projection, evaluation over long term period, and adequacy of applied discount rate, the economic assessment for back-end fuel cycle is different from each organizations or individuals. In this paper, the features and limitations of some noticeable economic evaluations were investigated and analysed to contribute for the public participation and back-end fuel cycle policy related researches. As a result of analysis, we found that the reprocess and recycling is more economical than direct disposal option, but the result includes high uncertainty that depends on the input parameters.
Development of a Computer Program for the Analysis Logistics of PWR Spent Fuels
Choi, Heui-Joo ; Cha, Jeong-Hun ; Choi, Jong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 147~154
It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.
Current Status and Characterization of CANDU Spent Fuel for Geological Disposal System Design
Cho, Dong-Keun ; Lee, Seung-Woo ; Cha, Jeong-Hun ; Choi, Jong-Won ; Lee, Yang ; Choi, Heui-Joo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 6, issue 2, 2008, Pages 155~162
Inventories to be disposed of, reference turnup, and source terms for CANDU spent fuel were evaluated for geological disposal system design. The historical and projected inventory by 2040 is expected to be 14,600 MtU under the condition of 30-year lifetime for unit 1 and 40-year lifetime for other units in Wolsong site. As a result of statistical analysis for discharge burnup of the spent fuels generated by 2007, average and stand deviation revealed 6,987 MWD/MtU and 1,167, respectively. From this result, the reference burnup was determined as 8,100 MWD/MtU which covers 84% of spent fuels in total. Source terms such as nuclide concentration for a long-term safety analysis, decay heat, thermo-mechanical analysis, and radiation intenity and spectrum was characterized by using ORIGEN-ARP containing conservativeness in the aspect of decay heat up to several thousand years. The results from this study will be useful for the design of storage and disposal facilities.