• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.351-357
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• 2003

Transuranic elements such as Pu, Am and Cm in synthetic solution of spent nuclear fuel samples were determined by electrodeposition followed by alpha-spectrometry after separation using anion exchange and extraction chromatography in order to determine the transuranic elements in radwaste samples from nuclear power plants. Plutonium was separated by 12M HC1-0.1M HI as an eluent on anion exchange column. As a second step Am and Cm were separated in a group by DTPA-Lactic acid as the eluent on HDEHP coated column. The nuclides of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ separated were determined by alpha-spectrometry after electrodeposition in 0.1M $NaHSo_4$-0.53M $Na_2SO_4$buffer solution as an electrolyte. The recovery yields of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ were 83.8%, 85.2% and 86.3%, respectively, from the synthetic solution containing uranium and non-radioactive metal elements.

• Journal of the Korean Geotechnical Society
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• v.34 no.1
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• pp.17-24
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• 2018

A buffer in a geological disposal system minimizes groundwater inflow from the surrounding rock and protects the disposed high-level waste (HLW) against any mechanical impact. As decay heat of a spent fuel causes temperature variation in the buffer that affects the mechanical performance of the system, an accurate estimation of the temperature variation is substantial. The temperature variation is affected by thermal and material properties of the system such as thermal conductivity, density and specific heat capacity of the buffer, and thus these factors should be properly included in the design of the system. In particular, as the thermal properties are variable depending on the density and water content of the buffer, consideration of the effects should be included in the analysis. Hence, in this study, a numerical model based on finite element method (FEM) which is able to consider the change of density and water content of the buffer was established. In addition, using the numerical model, a parametric study was conducted to investigate the effect of each thermal property on the temperature variation of the buffer.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.73-78
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• 2009

The decay heat of Cs and Sr contained in a LiCl waste salt, generated from an electrolytic reduction process in pyroprocessing of spent nuclear fuel, has been calculated. The calculation has been carried out under some assumptions that most of the LiCl waste is purified and recycled to main process, and the residual is fabricated to make a waste form. As a result, the decay heat from daughter nuclides such as Ba and Y seems to be maximum 4.6 times higher than that from their parent nuclides such as Cs and Sr. The thermal release from Cs and Sr in the LiCl waste is the maximum around the first one month, so an cooling system operation for some time at the beginning would be suggested to control a rapid increase in the temperature of the LiCl waste salt.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.84-92
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• 2023

As temporary storage facilities for spent nuclear fuels in domestic nuclear power plants are expected to be saturated, external intermediate storage facilities would be required in the future. Spent nuclear fuels are stored in metal canisters and then placed in a dry environment within concrete or metal casing for operation. In the United States, the dry storage method for spent nuclear fuels has been operated for an extended period. Based on the corrosion experiences of dry storage canisters in chloride environments, numerous studies have been conducted to reduce corrosion in welds. With the construction of intermediate storage facilities in Korea for spent nuclear fuels expected near coastal areas adjacent to nuclear power plants, there is a need for research on the corrosion occurrence of welds and mitigation methods for canisters in chloride environments. In this paper, we measured and compared the residual stresses in the Heat-Affected Zones (HAZ) after electron beam welding (EBW) and gas tungsten arc welding (GTAW) processes for candidate materials such as 304L, 316L, and duplex stainless steel(DSS). We investigated the possibility of microstructure control through the application of surface modification processes using friction stir processing (FSP). Corrosion tests on each welded specimen revealed a higher corrosion rate in EBW welds compared to GTAW. Furthermore, it was confirmed that corrosion resistance improved due to phase refinement and redistribution of precipitates when FSP was applied.

• Tunnel and Underground Space
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• v.34 no.2
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• pp.127-142
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• 2024

The robustness of a numerical method means that its computational performance is maintained under various modeling conditions. New numerical methods or codes need to be assessed for robustness through benchmark testing. The TOUGH-FLAC modeling approach has been applied to various fields such as subsurface carbon dioxide storage, geological disposal of spent nuclear fuel, and geothermal development both domestically and internationally, and the modeling validity has been examined by comparing the results with experimental measurements and other numerical codes. In the present study, a benchmark test of the TOUGH-FLAC approach was performed based on a coupled thermal-hydro-mechanical behavior problem with an analytical solution. The analytical solution is related to the temperature, pore water pressure, and mechanical behavior of a fully saturated porous medium that is subjected to a point heat source. The robustness of the TOUGH-FLAC approach was evaluated by comparing the analytical solution with the results of numerical simulation. Additionally, the effects of thermal-hydro-mechanical coupling terms, fluid phase change, and timestep on the computation of coupled behavior were investigated.

• Journal of Radiation Industry
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• v.18 no.2
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• pp.133-140
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• 2024

The Korean 2^nd basic plan for management of high-level radioactive waste presented a plan to manage spent nuclear fuel through dry storage facilities in NPP on-site. For the construction and operation of the facility, it is necessary to develop the monitoring system of the integrity of spent nuclear fuel before operation. NUREG-1536 recommends that the theoretical cask array, typically in the 2×10 array, should account for shadowing effect among the dry storage casks. The objective of this study was to evaluate neutron flux accounting for shadowing effect among dry storage casks. The neutron release rate was evaluated using ORIGEN based on the design basis fuel condition. And the simulation of dry storage casks and evaluation of the shadowing effect were performed using MCNP. Shadowing effect of other dry storage casks was the highest at the center of the dry storage facility of the 2×10 array compared with the outside of the cask. The shadowing effect of neutron flux on the surface among the metal casks was approximately 18% at point 1, 23% at point 2, and 43% at point 3. For the concrete casks, the shadowing effect of neutron flux on the surface was approximately 46% at point 1, 51% at point 2, and 52% at point 3. This means that correction is necessary to monitor the integrity of spent nuclear fuel in each dry storage cask through evaluation of shadowing effect. The results of this study will be used for comparative analysis of neutron measurement data from spent nuclear fuels in dry storage cask. Additionally, the neutron flux evaluation procedure used in this study could be used as the basic data of safety assessment of dry storage cask and development of safety guide.

• Economic and Environmental Geology
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• v.56 no.4
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• pp.421-433
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• 2023

The final disposal of spent nuclear fuel(SNF) from nuclear power plants takes place in a deep geological repository. The metal canister encasing the SNF is made of cast iron and copper, and is engineered to effectively isolate radioactive isotopes for a long period of time. The SNF is further shielded by a multi-barrier disposal system comprising both engineering and natural barriers. The deep disposal environment gradually changes to an anaerobic reducing environment. In this environment, sulfide is one of the most probable substances to induce corrosion of copper canister. Stress-corrosion cracking(SCC) triggered by sulfide can carry substantial implications for the integrity of the copper canister, potentially posing a significant threat to the long-term safety of the deep disposal repository. Sulfate can exist in various forms within the deep disposal environment or be introduced from the geosphere. Sulfate has the potential to be transformed into sulfide by sulfate-reducing bacteria(SRB), and this converted sulfide can contribute to the corrosion of the copper canister. Bentonite, which is considered as a potential material for buffering and backfilling, contains oxidized sulfate minerals such as gypsum(CaSO4). If there is sufficient space for microorganisms to thrive in the deep disposal environment and if electron donors such as organic carbon are adequately supplied, sulfate can be converted to sulfide through microbial activity. However, the majority of the sulfides generated in the deep disposal system or introduced from the geosphere will be intercepted by the buffer, with only a small amount reaching the metal canister. Pyrite, one of the potential sulfide minerals present in the deep disposal environment, can generate sulfates during the dissolution process, thereby contributing to the corrosion of the copper canister. However, the quantity of oxidation byproducts from pyrite is anticipated to be minimal due to its extremely low solubility. Moreover, the migration of these oxidized byproducts to the metal canister will be restricted by the low hydraulic conductivity of saturated bentonite. We have comprehensively analyzed and summarized key research cases related to the presence of sulfates, reduction processes, and the formation and behavior characteristics of sulfides and pyrite in the deep disposal environment. Our objective was to gain an understanding of the impact of sulfates and sulfides on the long-term safety of high-level radioactive waste disposal repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.321-332
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• 2017

In order to build a general model of a high-throughput uranium electrorefining process according to the electrode configuration, numerical analysis was conducted using the COMSOL Multiphysics V5.3 electrodeposition module with Ordinary Differential Equation (ODE) interfaces. The generated model was validated by comparing a current density-potential curve according to the distance between the anode and cathode and the electrode array, using a lab-scale (1kg U/day) multi-electrode electrorefiner made by the Korea Atomic Energy Research Institute (KAERI). The operating temperature was $500^{\circ}C$ and LiCl-KCl eutectic with 3.5wt% $UCl_3$ was used for molten salt. The efficiency of the uranium electrorefining apparatus was improved by lowering the cell potential as the distance between the electrodes decreased and the anode/cathode area ratio increased. This approach will be useful for constructing database for safety design of high throughput spent nuclear fuel electrorefiners.

• Magazine of the Korean Society of Agricultural Engineers
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• v.37 no.5
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• pp.62-72
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• 1995

공학분야에 널리 사용되고 있는 신뢰도 분석 방법 중에서 Monte Carlo simulation (MC), Mean-value First-Order Second-Moment Method(MFOSM), and Advanced First-Order Second-Moment(AFOSM) method들을 강물의 오염물질 농도와 수질기준치사이의 신뢰도 분석에 적용하였다. 미 환경 보건국에서 개발 보급한 QUAL2E를 이용하여 Mew Jersey에 위치한 Passaic강의 수질예측에서 4가지 주요인자(용존산소, 생물학적 산소요구량, 암모니아 그리고 조류)들이 정해진 수질기준치를 유지 할 수 있는 확률을 세가지 방법에 의해 추정한 후에 상호 비교하였다. MC방법에 의해 2,000회 simulation시켜서 그 결과가 시스템의 추계학적 성질을 잘 반영한 것으로 판단하여 비교기준으로 삼고 MFOSM과 AFOSM에 의해 추정한 결과와 비교하였다. MFOSM의 결과보다는 AFOSM의 결과가 전체적으로 MC의 결과에 더 근접하였으며, 이유는 AFOSM의 계산방법이 MFOSM의 선형근사로 인한 오차를 줄일 수 있었기 때문인 것으로 판단된다. MC방법의 결과와 다른 방법들의 결과사이의 차이가 입력 변수들이 평균값에서 멀어질 때가 많았는데 이는 MC의 경우 입력 변수들이 일정범위를 벗어나서 비현실적인 상황이면 model이 정지하는데, 다른 방법들은 simulation에 의한 것이 아니고 수학적인 계산에 의해서 신뢰도가 추정되기 때문에 이러한 상황이 반영될 수 없기 때문이다. 강물의 수질을 취급하는 공학적인 측면에서 보면, 이중에 가장 간편한 MFOSM이 많은 simulation이 필요한 MC나 계산방법이 상대적으로 복잡한 AFOSM에 비해 오차가 크지 않아서 이들을 대시하여 사용될 수 있다고 판단된다. 유래의 PAF가 분비된다는 것을 알 수 있었으며, 이러한 인자는 동결처리에서도 그 기능은 전혀 변하지 않는다고 본다. 이후에 있어서 mouse LIF의 첨가는 돼지의 수정란을 배반포 이후의 단계에까지 발달시킬 수 있었다. 있어서 더 적합한 것으로 판단되었다. 5. 개발된 모형은 논 관개의 물리적 측면과 관리목표 모두를 고려한 것으로 계산된 효율은 벼, 생육 각 단계에서의 효율 비교에 양호한 방법임을 알 수 있다.은 Sharpsburg 점질양토에 대한 S.C.S 한계허용치 10ton/ha/year 이내로 나타났다. 비처리구에서의 토양유실량은 평균 2.56ton/ha/year로 높게 나타난 반면 3개의 서로 다른 추리구인 비수구, 초생수로구 및 Bromegrass구에서는 각각 0.152, 0.192 및 0.290ton/ha/year로 낮은 결과를 가져왔다. 6. 평균 침전량에 대한 L.S.D. 검정 걸과 전시험구중 비처리구가 고도의 유의차를 나타낸 반면 비수구, 초생수로구 및 Bromegrass 목초구 간에는 아무런 유의차가 인정되지 않았다. 7. 농지보전 처리구인 배수구와 초생수로구는 비처리구에 비해 낮은 침두 유출량과 낮은 토양유실량을 나타내었다.구보다 14% 절감되는 것으로 나타났다.작용하는 것으로 사료된다.된다.정량 분석한 결과이다. 시편의 조성은 33.6 at% U, 66.4 at% O의 결과를 얻었다. 산화물 핵연료의 표면 관찰 및 정량 분석 시험시 시편 표면을 전도성 물질로 증착시키지 않고, Silver Paint 에 시편을 접착하는 방법으로도 만족한 시험 결과를 얻을 수 있었다.째, 회복기 중에 일어나는 입자들의 유입은 자기폭풍의 지속시간을 연장시키는 경향을 보이며 큰 자기폭풍일수록 현저했다. 주상에서 관측된 이러한 특성은 서브스톰 확장기 활동이 자기폭풍의 발달과 밀접한 관계가

• Journal of the Korean Electrochemical Society
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• v.18 no.4
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• pp.156-160
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• 2015

Experiments using a metal oxide of a non-nuclear material as a fuel are very useful to develop a new electrolytic reducer for pyroprocessing. In this study, the titanium oxides (TiO and $TiO_2$) were selected and investigated as the non-nuclear fuel for the electrolytic reduction. The immersion tests of TiO and $TiO_2$ in a molten 1.0 wt.% $Li_2O$-LiCl salt revealed that they have solubility of 156 and 2100 ppm, respectively. Then, the Ti metals were successfully produced after the separate electrolytic reduction of TiO and $TiO_2$ in a molten 1.0 wt.% $Li_2O$-LiCl salt. However, Ti was detected on the platinum anode used for the electrolytic reduction of $TiO_2$ unlike TiO due to the dissolution of $TiO_2$ into the salt.