• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.3
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• pp.199-206
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• 2013

Electrowinning process in pyroprocessing recovers U (uranium) and TRU (Trans Uranium) elements simultaneously from spent fuels using a liquid cadmium cathode (LCC). When the solubility limit of U deposits over 2.35wt% in Cd, U dendrites were formed on the LCC surface during the electrodeposition at $500^{\circ}C$. Due to the high surface area of dendritic U, the deposits were not submerged into the liquid cadmium pool but grow out of the LCC crucible. Since the U dendrites act as a solid cathode, it prevents the co-deposition of U and TRUs. In this study, the electrodeposition of U onto a LCC was carried out at 440 and $500^{\circ}C$ to compare the morphology and component of U deposits. The U deposits at $440^{\circ}C$ have a specific shape and were stacked regularly at the center of the LCC pool, while the U dendrites (i.e., ${\alpha}$-phase) at $500^{\circ}C$ were grow out of the LCC crucible. Through the microscopic observation and XRD analysis, the electrodeposits at $440^{\circ}C$, which have a round shape, were identified as an intermetallic compound such as $UCd_{11}$. It can be concluded that the LCC electrowinning operation at $440^{\circ}C$ achieves the co-recovery of U and TRU without the formation of U dendrites.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.291-291
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• 2009

한국원자력연구원의 파이로 실험 시설인 ACPF (ACP Facility)에는 공학규모 전해환원 반응기가 설치되어 공정 대용량화를 위한 연구가 수행되고 있다 본 연구에서는 전해환원 공정의 Scale-up을 위해 기존 반응기를 개선하여 전해환원 실험을 수행한 결과를 담고 있다. 장치의 대형화 빛 원격운전성 향상을 위해 기존의 전해환원 반응기의 상부 플랜지는 보다 간단하게 정리되었으며 염 이송에 의한 고온 조건 노출 시간을 줄임과 동시에 염 재사용을 목적으로 상부 플랜지는 이중으로 설계되었다. 따라서, 반응 종료후 전극이 설치된 상부 플랜지를 들어 올림으로서 반응기를 불활성 분위기로 유지하는 동시에 전해환원 금속전환체를 회수 할 수 있도록 반응기가 제작되었다. 또한, 새로운 반응기는 용융염 내의 강제 유동을 위해 아르곤 버블링이 가능하도록 설계 제작되었다. 새로 제작 설치된 전해환원 반응기를 사용하여 산화물 분말을 혼합하여 준비한 모의 사용후핵연료를 사용하여 전해환원 실험을 수행하였다. 그 결과, 산화물이 충진된 음극의 전영역에서 고루 96% 이상의 높은 금속전환율을 얻었으며 시간에 따라 선택된 FP들의 용융염 내 거동을 측정하였다. 실리더 형태의 음극에서 Cs, Sr 등의 원소들이 용융염으로 시간에 따라 용출되는 것을 확인하였으며 동시에 반응기 재질인 Fe 등도 일부 용융염에서 검출되었다. 아르곤 버블링에 의한 강제 유동은 전압 및 전류 거동에는 큰 영향을 미치지 못하였으나 염의 휘발량을 증가시켜 영조성올 변화시키는 것으로 측정되었다. ACPF의 전해환원 실험결과를 바탕으로 반응기를 상부 기체상과 하부 액체상으로 나누어 전산모사를 수행하였다 상부 기체상은 유입되는 아르곤 기체와 발생되는 산소기체의 흐름을 모사하는 결과를 얻었으며 온도 및 산소의 분압을 계산하였다. 하부 액체상에서는 전기장을 모사하여 전류 밀도 등을 3차원으로 모사하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.1
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• pp.17-23
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• 2009

Dihexyloctanamide(DHOA) was used as an extractant or phase modifier with the diamide extractants in a solvent extraction process for a radioactive liquid waste treatment. The degradation compounds of the DHOA extractant, irradiated with $^{60}Co$ gamma ray, were octanoic acid and dihexylamine which are identified by a Fourier transform infrared(FT-IR) and gas chromatograph/mass spectrometer(GC/MS) analysis, and determined by the GC/MS with selected ion monitoring(SIM) mode. Retention behavior of octanoic acid, tridecane (internal standard) and dihexylamine in total ion chromatogram (TIC) were 8.65 min., 9.79 min., and 10.27 min., respectively. With increasing the absorbed dose of the $\gamma$-ray irradiated DHOA, the concentration of octanoic acid was decreased and that of dihexylamine was increased.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.3
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• pp.219-230
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• 2004

Most of the low-level liquid radioactive wastes generated from PWR plants are classified into high or low total suspended solid(HTDS or LTDS), and into radiochemical and radioactive laundry waste. Although the evaporation process has a high decontami- nation ability, it has several problems such as corrosion, foam, and congestion. A new liquid waste disposal process using the ion-exchange demineralizer(IED), instead of the current evaporation process, has been introduced into the Yonggwang NPP #5 and 6. These two methods have been compared to understand the differences in this study. Aspects compared here were the released radioactivity amount of the liquid radioactive wastes, the dose of off-site residents, the decontamination factor, and the amount of the solid radioactive wastes. The IED system is designed to discharge higher radioactivity about 20% than the evaporating system, and the actual radioactivity released from the evaporating and IED system were 0.473mCi and 1.098mCi, respectively. The radioactivity released from the IED was 2.32 times higher than that of the evaporating system. The dose of off-site residents was $2.97{\times}10^{-6}$mSv for the evaporating system, and $6.47{\times}10^{-6}$mSv for IED. The decontamination factor(DF) of the evaporator is, in most cases, far lower than the lower limits of detection(LLD) with the Ge-Li detector. Due to the low concentration of the liquid wastes collected from the liquid waste system, the decontamination factor of IED is very low. Since there is not enough data on the amount of solid radioactive wastes generated by the evaporation system, the comparison on these two systems has been conducted on the basis of the design, and the comparison result was that the evaporating system generated more wastes about 40% than IED.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.83-89
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• 2003

A temperature- and pressure-reducing process is utilized to handle the spent fuel assembly in the post-irradiation examination facility. This process includes three separated unit processes. First one is the decontamination process to clean the spent fuel assembly casks. The second process is the temperature-reducing process to reduce the temperature elevated by decay process in the spent fuel assembly. The third process is the filtration process to remove insoluble particles existed in the casks using filters. Up-to-date technologies as well as practical theories related to the temperature- and pressure-reducing process is reviewed in this report. The test-operation process for various tests and the test results of the temperature- and pressure-reducing process for J-44 and K-23 spent fuel assemblies are also described in detail. This report must be effectively used for the normal operation of the facility with the awareness of unprecedented problems which could occur by continuing operation of the PIE facility.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.165-172
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• 2018

As an approach for estimation of the droplet size in the molten salt-liquid metal extraction process, a droplet formation experiment at room temperature was conducted to evaluate the applicability of the Scheele-Meister model with water-mercury system as a surrogate that is similar to the molten salt-liquid metal system. In the experiment, droplets were formed through the nozzle and the droplet size was measured using a digital camera and image analysis software. As nozzles, commercially available needles with inner diameters (ID) of 0.018 cm and 0.025 cm and self-fabricated nozzles with 3-holes (ID: 0.0135 cm), 4-holes (ID: 0.0135 cm), and 2-holes (ID: 0.0148 cm) were used. The mercury penetration lengths in the nozzles were 1.3 cm for the needles and 0.5 cm for the self-fabricated nozzles. The droplets formed from each nozzle maintained stable spherical shape up to 20 cm below the nozzle. The droplet size measurements were within a 10% error range when compared to the Scheele-Meister model estimates. The experimental results show that the Scheele-Meister model for droplet size estimation can be applied to nozzles that stably form droplets in a water-mercury system.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.419-424
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• 2004

Various radiation counters have been using to determine radioactivity of radwastes for disposal. A radiation counting system was set up using a radiation detector chosen in this study and its stability was investigated through the periodic determination of background and counting efficiencies in accordance with a quality control program to increase the confidence level. The average background level for the $\gamma$-spectrometer was 1.59 cps and the average counting level for the standard sample was 45248 Ops within $2{\sigma}$ confidence levels. The average alpha background level for the low background ${\alpha}{\beta}$ counting system was 0.31 cpm and the efficiency for alpha counting was 34.38%. The average beta background level for the ${\alpha}{\beta}$ counting system was 1,30 cpm and the efficiency for beta counting was 46.5%, The background level in the region of 3H and 14C for the liquid scintillation counting system was 2.52 and 3.31 cpm and the efficiency for alpha counting was 58.5 and 95.6%, respectively. The minimum detectable activity for the$\gamma$-spectrometer was found to be 3.2 Bq/$m\ell$ and 3.8 Bq/$m\ell$ for the liquid scintillation counter, and 20.5 and 23.0 Bq/$m\ell$, respectively for the $\alpha$ and $\beta$ counting system.

• Clean Technology
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• v.14 no.2
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• pp.110-116
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• 2008

The radioactive wastes generated from the nuclear industry can be divided into the forms of solid, liquid, or gas. Radioactive methyl iodide, a gaseous radioactive waste, is absorbed by activated carbon with 5 wt% of Trietylenediamine (1,4-diazania-bicycle[2.2.2]octane, TEDA) impregnated on the surface. Methyl Iodide ($CH_3I$) is combined chemically with TEDA (the final product : I-TEDA). To recycle radioactive activated carbon, removal of I-TEDA from activated carbon is needed. A wet method for recycling impregnated active carbon was developed to remove radioactive I-TEDA using an acetonitrile solution, which produces lots of secondary wastes. We suggest the removal of I-TEDA by supercritical carbon dioxide with co-solvents. In this experiment, we used a quartz crystal microbalance (QCM) for measuring the removal rate of the I-TEDA. From the experimental results, methanol was found to be the optimum co-solvent, and the optimum conditions such as temperature, pressure, and co-solvent flow rate were obtained. Possibility of using supercritical fluid in the removal of I-TEDA from radioactive activated carbon was also discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.345-349
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• 1997

고준위 방사성 액체폐기물에서 얻어지는 백금족 금속(Pd, Rh, Ru) 들의 분리 및 정제방법으로 강염기성 음이온교환수지를 사용하여본 결과 상용 수지중에서 Dowex 1 $\times$ 8 이 IRN-78 에 비하여 저 농도의 질산 농도에서 Pd(II) 의 분리 및 정제시 우수한 흡착성을 보여 주었으며 Rh(III) 의 흡착은 Pd(II) 의 것보다 훨씬 낮은 값을 보여 주었다. 이 수지들의 백금족 금속에 대한 흡착성을 문헌에 보고된 실험 결과들과 비교 검토하여 본 바 이온 그룹으로 3급 및 4급 Benzimidazole을 가지는 수지에 비하여 훨씬 낮은 값을 나타내었다. 따라서 실용성이 큰 강염기성 음이온수지로서는 Benzimidazole과 같은 혼합 아민 그룹을 지닌 수지가 가장 접합할 것으로 전망되었다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.329-334
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• 2006

The depth distributions of total $^3H$ and $^{14}C$ activities were characterized for the activated shielding concrete from a decommissioning of KRR-2 using the commercially available tube furnace and a liquid scintillation counter. The correlation of measurement results between $^3H,\;^{14}C$ and gammer emitter was evaluated to apply for estimating radionuclide inventory of the concrete waste generated from decommissioning KRR-2. The detection limits for $^3H$ and $^{14}C$ are 0.048 and 0.028 Bq/g respectively. The specific activities of the $^3H$ and $^{14}C$ tend to decrease exponentially as the depth of the concrete becomes deeper from the surface. In addition, the $^3H$ and $^{14}C$ activities were in good correlation with the $^{60}CO$ activities analysed for the shielding concrete of KRR-2.