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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
for Alpha Spectrometry and Application to Spent Nuclear Fuel Samples
Joe Kih-Soo ; Kim Jung-Suck ; Han Sun-Ho ; Park Yeong-Jai ; Kim Won-Ho ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 95~102
Alpha spectrometry was studied for the determination of
in spent nuclear fuel samples. The optimum condition for the electrodeposition of
was obtained as follows : for
hour of deposition time, at the current intensity of
A and at sodium sulfate electrolyte without organic additive. The deposition yield and its reproducibility on
was decreased as the amount of
decreased from 4.16 Bq down to 0.0264 Bq(1ng). The recovery yield of
determined by alpha spectrometry after separation in synthetic solution was
(n=4). The contents of
in spent nuclear fuel samples were determined and the result showed an agreement within 10% of a difference between the measurement and the calculation.
An Experimental Research on Uniform Corrosion of Inconel 600 and 690 Tubing Material
Yeom Yu-Sun ; Hwang Jung-Lae ; Jun In-Sub ; Kim Soong-Pyung ; Yoon Jang-Hee ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 103~116
By executing corrosion experiment on Inconel 600, 690 used to material of S/G tube in domestic NPP, this paper show estimation of amount of product such as Co-58, Co-60, Cr-51, Mn-54, Fe-59 which are main exposure cause to the workers in NPP. Therefore, Making the 12 samples consisted of Inconel 600, 690, whole corrosion experiment was carried out for 60 days(each pH by 20 days). The conditions of those tests were similar or more harsh than actual conditions of domestic NPP. The Glow Discharge Spectrometer(GDS) was used for quantitative analysis of results. The results of using GDS, the Inconel 600 corrodes more than Inconel 690 at pH 7 and pH 9. However, it is observed that Inconel 690 corrodes more than Inconel 600 at pH 4. Those results is estimated that test sections had the effect of transient. The long terms of experiment is required to minimize and solve the problem.
Coupled T-H-M Processes Calculations in KENTEX Facility Used for Validation Test of a HLW Disposal System
Park Jeong-Hwa ; Lee Jae-Owan ; Kwon Sang-Ki ; Cho Won-Jin ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 117~131
A coupled T-H-M(Thermo-Hydro-Mechanical) analysis was carried out for KENTEX (KAERI Engineering-scale T-H-M Experiment for Engineered Barrier System), which is a facility for validating the coupled T-H-M behavior in the engineered barrier system of the Korean reference HLW(high-level waste) disposal system. The changes of temperature, water saturation, and stress were estimated based on the coupled T-H-M analysis, and the influence of the types of mechanical constitutive material laws was investigated by using elastic model, poroelastic model, and poroelastic-plastic model. The analysis was done using ABAQUS, which is a commercial finite element code for general purposes. From the analysis, it was observed that the temperature in the bentonite increased sharply for a couple of days after heating the heater and then slowly increased to a constant value. The temperatures at all locations were nearly at a steady state after about 37.5 days. In the steady state, the temperature was maintained at
at the interface between the heater and the bentonite and at about
at the interface between the bentonite and the confining cylinder. The variation of the water saturation with time in bentonite was almost same independent of the material laws used in the coupled T-H-M processes. By comparing the saturation change of T-H-M and that of H-M(Hydro-Mechanical) processes using elastic and poroelastic material mod31 respectively, it was found that the degree of saturation near the heater from T-H-M calculation was higher than that from the coupled H-M calculation mainly because of the thermal flux, which seemed to speed up the saturation. The stresses in three cases with different material laws were increased with time. By comparing the stress change in H-M calculation using poroelasetic and poroelasetic-plastic model, it was possible to conclude that the influence of saturation on the stress change is higher than the influence of temperature. It is, therefore, recommended to use a material law, which can model the elastic-plastic behavior of buffer, since the coupled T-H-M processes in buffer is affected by the variation of void ratio, thermal expansion, as well as swelling pressure.
Improvement of Removal Characteristics of Uranium by the Immobilization of Diphosil Powder onto Alginate Bed
Kim Kil-Jeong ; Shon Jong-Sik ; Hong Kwon-Pyo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 133~138
Chemical wastes containing small amounts of uranium can not be disposed of them after treatment as an industrial waste, because the uranium concentration in the final dry cake exceeds the exemption level. Especially for the removal of uranium in this study, the method for immobilizing Diphosil powder within alginate beads is adopted to make a bead form from a powdered resin. Sodium alginate bead itself showed a capability to uptake uranium to above 60%, but the value was decreased to below 30% after equilibrium. The adsorption rate of uranium increased with the increasing content of Diphosil in the sodium alginate bead. Diphosil resin itself showed very fast uptake of uranium from early stages, and then the rates were leveled off. Diphosil bead showed an improved capability to uptake uranium considering the pure Diphosil content in the composite bead, and provide a considerable potential for further applications of a continuous process by using Diphosil as a bead form.
A Structural Analytic Evaluation of a Connote Pad In a Spent Fuel Dry Storage Cask
Kim Dong-Hak ; Seo Ki-Seog ; Lee Ju-Chan ; Lee Yeon-Do ; Cho Chun-Hyung ; Lee Dae-Ki ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 139~152
A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a lee drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.
Bio-Denitrification of the Nitrate Waste Solution from the Lagoon Sludge in a Batch Fermenter
Oh Jong-Hyeok ; Lee O-Mi ; Hwang Doo-Seong ; Choi Yun-Dong ; Hwang Sung-Tae ; Jo Byung-Real ; Park Jin-Ho ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 153~159
It is a serious task to the decommissioning of the uranium conversion plant that the demolition of the lagoon sludge. The main component of the sludge is ammonium nitrate and that is the very explosive material. Therefore, the bio-denitrification is a attractive process to remove the nitrate. In this work, some process variables was tested such as incubation temperature, nitrate concentration, electron donor, C/N ratio, seeding ratio, and pH with an anaerobic bacteria as Pseudomonas halodenitrificans. The results would be used as basic data to the continuous bio-denitrification process.
Development and Performance Evaluation of a Filtration Equipment to Reuse PFC Waste Solution Generated on PFC Decontamination
Kim Gye-Nam ; Jeong Cheol-Jin ; 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 2, 2006, Pages 161~170
PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered on the inner surface of hot cell and surface of equipment in hot cell. It was necessary to develop a filtration equipment to reuse the PFC waste solution generated on PFC decontamination due to the high cost of PFC solution and for minimization of the volume of second waste solution. The filtration equipment was developed to remove hot particulate in PFC waste solution. It was made suitable size and weight in consideration of hot cell gate and crane. And it has wheels for easy movement. Flux of the filtration equipment decreased with particulate concentration increase. It consists of pre-filter(
) and final-filter(
) for protection of the flux decrease along filtration time. It treatment capacity of waste solution is 0.2 L/min.
A Framework of Decommissioning Cost Estimation for Nuclear Research Facilities
Jeong Kwan-Seong ; Lee Dong-Gyu ; Lee Kune-Woo ; Oh Won-Zin ; Jung Chong-Hun ; Park Jin-Ho ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 171~178
Decommissioning cost estimation is a very important technique in designing and planning of nuclear facilities' decommissioning. Decommissioning cost estimation should be made according to the phases of decommissioning activities and installed components of nuclear facilities. In this paper, the basic framework necessary for decommissioning cost estimation is completed so that it could be used as a technique for decommissioning costs estimation by specifying cost items and group components and unit cost factors on which work time is calculated. Also, factors to be considered for decommissioning cost estimation of major activities and tasks are reviewed. Afterwards, these techniques will be utilized as a basic technology to establish methodology of decommissioning cost estimation and evaluation.
Alternative Method for the Treatment of Chemical Wastes Containing Uranium
Kim Kil-Jeong ; Shon Jong-Sik ; Hong Kwon-Pyo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 179~186
Chemical wastes are generated from nuclear facilities and R&D laboratories, but the uranium concentration in the final dried cake is evaluated into 11.2 Bq/g, which exceeds the exemption level of 10 Bq/g for each U isotopes, so the cake is categorized into a radioactive waste. Acid dissolution was applied to extract uranium from the waste sludge, and uranium adsorption on the dissolved solution was experimented by using IRN-77 and Diphosil bead. A large amount of resin was required to get above 80% of uranium removal, which was found to be due to a large amount of metal ions simultaneously dissolved from the precipitates with uranium. As an alternative method, acid dissolution is applied to the dewatered wet cake of the sludge, and the natural evaporation method is adopted for the dissolved solution. The uranium concentration of the dissolved solution was estimated to be 6.97E-01 Bq/ml, and the specific activity of the final waste sheets is evaluated to be 4.3 Bq/g. These results lead to the suggestion that the application of acid dissolution to the wet cake and the natural evaporation for the dissolved solution is an effective treatment method for chemical wastes containing uranium.
Low and Intermediate Level Radioactive Waste Certification Program Plan
Ahn Sum-Jin ; Kim Tae-Kook ; Lee Young-Hee ; Kang Ill-Sik ; Shon Jong-Sik ; Hong Kwon-Pyo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 187~195
The regulation for the low and intermediate level radioactive waste to be transferred to the disposal facility, recently revised, require that radioactive waste generators should set up waste certification program to verify the radioactive waste conform to the waste acceptance criteria(WAC) before disposal. The radioactive waste disposal facility, scheduled to be constructed in Korea, will institute WAC for the wastes to be transferred to the facility. This WAC is expected to compose of the requirements for the radiological characterization, physical and chemical characterization, physical/chemical restriction, prohibited item, packaging, identification, labeling, and documentation. For the compliance with this regulation, The radioactive waste generators should verify that the waste meet WAC through performance of the waste certification program and are responsible for handing in all the certification documents to the disposal facility. This waste certification program plan was set up as a preliminary program for the certification of radioactive waste generated in Korea Atomic Energy Research Institute (KAERI) and should be further revised until preparation of WAC by disposal agent.
Engineering-scale Validation Test for the T-H-M Behaviors of a HLW Disposal System
Lee Jae-Owan ; Park Jeong-Hwa ; Cho Won-Jin ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 4, issue 2, 2006, Pages 197~207
The engineering performance of a high level waste repository is significantly dependent upon the T-H-M behavior in the engineered barrier system. An engineering-scale test facility (KENTEX) was set up to validate the T-H-M behaviors in the buffer of a reference disposal system developed in the 2002. The validation tests started on May 31, 2005 and is now in progress. The KENTEX facility and validation test programme are introduced, and pre-operation calculations are also presented to give information on the sensitive location of sensors and operational conditions. This test will provide information (e.g., large-scale apparatus, sensors, monitoring system etc.) needed for 'in-situ' tests, make the validation of a T-H-M model for the T-H-M performance assessment of the reference disposal system, and demonstrate the engineering feasibility of fabricating and emplacing the buffer of a repository.