• Proceedings of the Korean Institute of Resources Recycling Conference
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• 2005.10a
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• pp.330-334
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• 2005

본 연구에서는 유해 중금속을 다량 함유하고 있는 산업폐기물의 고화 처리에 사용되는 칼슘설포알루미네이트(4CaO $3Al_2O_3\;SO_4$ 이후부터 CSA로 기입)의 제조를 위해 철강부산물인 압연 슬러지를 활용하여 그 특성에 대해 조사하였다. 본 연구에서는 철강 부산물인 압연슬러지 외에 석회석 미분물, 인산부산 석고를 혼합하여 칼슘알루미네이트 상을 합성하였다. 합성 결과 소성온도 $1250^{\circ}C$에서부터 CSA가 합성되었고, 이와 함께 칼슘실리케이트$(2CaO\;SiO_2)$와 칼슘알루미노페라이트($4CaO\;Al_2O_3\;Fe_2O_3$)도 함께 합성되었다. CSA 합성에 미치는 중금속 영향을 관찰한 결과 원료의 중금속이 CSA 합성 온도를 낮추는 효과가 있는 것으로 나타났다. CSA를 이용한 철강산업 폐기물의 중금속 고용 처리 연구에서도 본 실험에서 합성된 CSA가 폐기물의 중금속 고화 처리에 효과가 있는 것으로 나타났다.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2022.11a
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• pp.167-168
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• 2022

Recently, starting with the permanent suspension of Gori 1 in Korea, the importance of the disposal of concrete structures in nuclear power plants has emerged, and environmental and safety are required to be proved accordingly. Safe radioactive waste disposal technology that immobilizes harmful radioactive elements, which are by-products of nuclear power, inside a solid matrix and recycling measures are needed to secure an efficient waste disposal space. This study was conducted to confirm whether recycled cement generated in the process of radioactive concrete treatment can be used as a solidifying material for radioactive waste treatment. In order to simulate the concrete exposed to radiation, aqueous solutions of Di-water, CsCl 1M, and CoCl₂ 1M were used as blending water at W/B 0.5. Tricalcium phosphate and Prussian blue were substituted with 5 wt.% based on the weight of recycled cement as a binder to improve solidification performance, and their hydration characteristic was analyzed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.177-177
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• 2009

사용후 핵연료의 건식처리 시 핵연료 다발을 절단하여 voloxidation 즉 휘발산화처리를 하면 고온에 의해 분리가 가능한 핵분열생성물의 분리와 우라늄의 산화에 의한 부피팽창으로 핵연료가 쪼개져서 입도가 작아지고 또한 핵연료가 피복재에서 쉽게 박리되게 된다. 그 결과 폐기물 처리 시에 발열핵종으로 폐기물의 저준위화시에 분리가 요망되는 Cs-137이 분리되는 장점이 있어 습식 재처리에 있어서도 바람직하다. 건식처리에 있어서는 voloxidation 으로 처리된 피복재에는 금속 지르코늄에 불순물로 함유된 우라늄의 의한 방사화 생성물과 피복재 표변에 부착/침투한 방사화 생성물이 방사능을 갖게 된다. 이러한 부착된 TRU 잔류물은 통상 1% 미만으로 알파핵종의 방사능이 원자로에서 배출시에는 고준위 기준치의 약 100배 수준이었다가 30년 냉각후에는 약 1/10 수준으로 저준위화 된다. 지르코늄 금속중에 불순물로 함유된 우라늄의 방사화로 생기는 방사능은 고준위 기준치의 10% 를 넘지 않아서 피복재의 저준위화시에 고려할 필요가 없다. 발생열은 방출시에 고준위 기준치의 약 30 배 수준에서 5년 냉각후에는 기준치 미만이 되며 30년후에는 1/8000 정도로 저준위화 된다. 사용후 핵연료를 습시처리시에 발생하는 고준위 폐기물 중 약 1/4 가 피복재 (hull) 임을 고려하면 피복재의 저준위화는 사용후 연료의 건식처리에 있어서도 필수적인 과정이다. 특히 미국의 고준위 폐기물 처분장 Yucca Mt.의 포기와 우리의 고준위 폐기불 처분장이 공론화되는 싯점에서 저준위화는 매우 필요한 기술이다. 피복재는 방사성 물질의 침투두께가 0.01mm 미만이 대부분으로 저준위화에는 표면제염에 의한 저준위화가 주로 연구되어왔다. 표면제염에 의한 저준화는 이온 빔, laser에 의한 방법, dry ice 분사에 의한 방법이 시도되었다. 염소기체를 이용하여 지르코늄의 산화막을 제거하고자 하였으나 이 산화막이 안정적이어서 표변의 연마, 아크릴 칼의 사용, 표면을 눌러서 처리하는 등 전처리하여서 염소기체 반응에 의한 표면제거 실험이 가장 효과적임이 실험적 결과이었다. 이러한 전처리로 방사능을 1/100 수준으로 낮춘다고 하더라도 지르코늄 금속중에 불순물로 함유된 우라늄의 방사화에 의해 중저준위 폐기물의 범주에서 벗어나지 않으므로 재활용에는 제한이 있다. 또한 전처리(표면제염)하여 분리되는 고준위는 다른 고준위 염폐기물과 함께 처리하여 발열 핵종을 제거하면 중저준위화가 가능하다. 저준위화 된 hull폐기물에는 지르코늄 금속에 불순물로서 함유되어있는 우라늄에 의한 방사능을 갖는데 이들의 제거나 분리는 지르코늄 합금 피복재 원료물질에 불순물로 함유하는 우라늄의 함량을 낮추는 것과 유사한 문제이다. 현재까지 지르코늄합금 피복재에 우라늄이 불순물로 함유된 것을 사용함으로 원자로내에서 방사화되어서 방사능을 갖게 되는 것은 피할 수가 없다. 따라서 저준위화 처리된 피복재는 장기 보관으로 방사능을 감쇠시켜서 재활용하도록 한다. 처리 방법으로는 초고압 압축저장, 시멘트 고화, 합성암석에 의한 고화법 등으로 장기간 보관 후에 금속으로서 재활용한다.

• Economic and Environmental Geology
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• v.39 no.2 s.177
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• pp.173-180
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• 2006

The synthesis of polyphase waste form, which is an immobilization matrix fur the high level radioactive wastes, was performed with the mixed composition of garnet and spinel $(Gd_3Fe_5O_{12}+(Ni_xMn_{1-x})(Fe_yCr_{1-y})_2O_4)$ in the range of 1200 to $1400^{\circ}C$. The phases synthesized from all stoichiometric compositions were garnet, perovskite, and spinel. Especially, garnet was synthesized only in the composition of the highest content of Fe(y=0.9), whereas it was not synthesized in other compositions. This result indicated that the content of Fe was closely related to the formation of garnet. The composition of garnet revealed that the content of Gd was exceeded and that of Fe was depleted. Preferential distribution of elements in the phases can be attributed to the nonstoichiometric composition of garnet.

• Nuclear Engineering and Technology
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• v.20 no.2
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• pp.120-131
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• 1988

A souce term model describes mathematically the source of radionuclides as they begin slow migration and decay in deep groundwater. Various source term models based on mass-transfer analysis and measurement-based source term models are reviewed. Ganerally, two processes are involved in leaching or dissolution: (1) chemical reactions and (2) mass transfer by diffusion. The chemical reaction controls the dissolution rates only during the early stage of exposure to groundwater. The exterior-field mass transfer may control the long term dissolution rates from the waste solid in a geologic repository. Mass-transfer analyses re3y on detailed and careful application of the governing equations that describe the mechanistic processes of transport of material between and within phases. If used correctly, source term models based on mass-transfer theory are valuable and necessary tools for developing reliable predictions.

• Journal of Korean Society of Environmental Engineers
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• v.30 no.3
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• pp.278-285
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• 2008

This research was performed to evaluate the landfilling method of organic sludge from mix of pre-treated organic sludge (OS) and municipal solid waste(MSW). Organic sludges were dried, composted, and solidified as pre-treatment and the OS and MSW were mixed in ratios of 2 to 8 and 4 to 6. Approximately 1,800$\sim$2,500 L of landfill gas(LFG) was generated in the lysimeter with solidified-OS, which was higher than 1,150$\sim$1,650 L of the dried- and composted- ones. Maximum H$_2$S concentration was found in the following order : Composted-20(80 ppmv) > Composted-40(55 ppmv) > Dried-20(30 ppmv) > Dried-40 $\fallingdotseq$ Solidified-20 $\fallingdotseq$ Solidified-40 (20 ppmv). BOD$_5$ at initial leachate generation period was 38,000 mg/L for Composted-40, 28,000 mg/L for Dried-40, 26,000 mg/L for Dried-20, 21,000 mg/L for Composted-20 and Solidified-40, and Solidified-20 for 17,000 mg/L. In the final period of experiment, BOD$_5$ was low as 300$\sim$500 mg/L in the lysimeter with solidified-OS and MSW and showed 2,000$\sim$3,500 mg/L in dried- and composted- ones. As the results, landfilling by mix of solidified-OS and MSW was evaluated as the most appropriate method for biodegradable organics. Direct landfilling of OS is permitted for landfill site with CDM facility. Therefore, mixed landfilling of solidified-OS and MSW should be considered for much more LFG generation as methane.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.227-234
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• 2006

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 $10g/m^2/day$ 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.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.197-202
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• 1988

A finite slab approximation method was developed to predict the long-term teachability. It is based on the assumption that the diffusional characteristics of radionuclides in a waste matrix are not dependent on matrix geometry but dependent on volume to surface ratio V/S) and diffusion coefficient. Consequently it can be expressed as the solution of the equations obtained from a finite slab with an equal V/S ratio (imaginary diffusion length). The calculational results by the finite slab approximation method have been compared with the results obtained for finite cylinder and sphere with corresponding diffusional analysis. The results of this simple model have showed a good agreement and presented a general applicability for the long-term prediction of the radionuclide leaching behavior.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.302-307
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• 1989

Nuclide release through penetrations in radioactive waste container is analyzed. Penetrations may result from corrosion or cracking and may be through the container material or through deposits of corrosion products. The analysis deals with the resultant nuclide release, but not with the way these penetrations occur. Numerical illustrations show that mass transport from multiple holes can be significant and may approach the mass transfer rate calculated from bare waste forms. Although partially-failed containers may present an important long-term barrier to release of radionuclides, numerous small holes on a container surface have the potential of bypassing the effectiveness of these barriers.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.215-219
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• 2003

The technology to fix a molten LiCl waste, which would be generated from the process to convert spent fuel to metal, into zeolite and then make a final waste form is doing developed. The XRD results of salt-loaded zeolites with different mixing ratios showed that all zeolites transformed from zeolite A type into Li-A type, or also Sodalite type as a minor phase for some conditions. The optimum LiCl-to-zeolite ratio to bring a minimum free salt was 1.0 when the molten LiCl waste contained Cs and Sr.