• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.194-195
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• 2005

방사선관리구역 내에서 발생되는 많은 양(발전소별 연간 약$100{\sim}700$개 정도)의 사용 후 공기정화 필터는 고체 폐기물 드럼의 단면적 보다 넓어 별도로 필터를 압축하거나 분해 작업 없이는 폐기물로 직접 처리가 곤란하다. 처리 시 많은 양의 분진이 발생하여 작업자의 내부피폭 가능성 및 많은 양의 고체 폐기물이 발생할 수 있는 잠재성이 있어 시료분석 결과 오염된 필터 내지는 바로 드럼 처리하여 내부 피폭 가능성을 미연에 방지하고 필터 프레임은 재사용을 유도하여 폐기물 저감화, 작업환경 개선 및 경제적인 이익을 창출할 수 있다. 영광 3발전소와 울진 3발전소의 경우 타 발전소에 비해 방사선관리구역 내 공기정화 처리기의 설치수량이 많아(약 700개/년) 공기정화필터가 매년 다수 발생되고 있으며, 이를 전량 드럼 처리 시 고체폐기물 드럼이 더 발생하게 되어 영구처분비용의 증가를 초래하게 된다. 발전소 전체적으로는 약 3,500개/년의 폐필터가 발생되고 있다. 이렇게 발생되는 공기정화필터의 프레임을 재사용함으로써 그 효과는 1) 알루미늄을 포함한 유리섬유를 드럼처리 시 고체방사성 폐기물드럼 생성량 감소 2) 프레임 재사용으로 인한 예산절감 효과 3) 폐필터 분해작업 시 분진에 의한 작업자 체내${\cdot}$외 피폭방지와 작업장 오염 확산 방지 및 환경 개선 4) 작업시간 단축 및 소요인력 감소 효과를 볼 수 있다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.247-254
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• 2012

Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating & Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of $SiO_2$ and $B_2O_3$ for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.109-116
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• 2003

The amount of radioactive waste is expected to be increased continuously because of the rapid growth of the domestic nuclear industry, full power operation of the HANARO reactor and the increased research activities of the nuclear fuel cycle. Accordingly the efforts are focused to achieve the handling of radioactive waste in safe and reduce the volume of radioactive waste. The PIEF is carrying out the PIE (post irradiation examination) of spent fuel rods related to the identification of cause defect and evaluation of integration safety. This study describes the technologies and experiences of compaction, shredding and cutting of the solid radioactive waste used in the PIE. The quantity of the high level waste was reduced by 1/12 using the 100-ton compressor installed in hot-cell. Also middle and low level waste was reduced by 1/8 using the 60-ton compressor installed in intervention area. Plastic drums were shredded by crusher to be compacted in the ratio of 1/5, used filters in the ratio of 1/6 and the number of drum is also reduced by cutting procedure for the non-volatile materials such as metal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.59-68
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• 2014

The public's access to the disposal facilities should be restricted during the institutional control period. Even after the institutional control period, disposal facilities should be designed to protect radiologically against inadvertent human intruders. This study is to assess the effective dose equivalent to the inadvertent intruder after the institutional control period thorough the GENII. The disposal unit was allocated with different kind of radioactive waste and the effects of the radiation dose to inadvertent intruder were evaluated in accordance with the institutional control period. As a result, even though there is no institutional control period, all were satisfied with the regulatory guide, except for the disposal unit with only spent filter. However, the disposal unit with only spent filter was satisfied with the regulatory guide after the institutional control period of 300 years. But the disposal unit with spent filter mixed with dry active waste could shorten the institutional control period. So the institutional control period can be reduced through the mixing the other waste with spent filter in disposal unit. Therefore, establishing an appropriate plan for the disposal unit with spent filter and other radioactive waste will be effective for radiological safety and reduction of the institutional control period, rather than increasing the institutional control period and spending costs for the maintenance and conservation for the disposal unit with only spent filter.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.279-298
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• 2019

This study comprehensively addresses recent progress at KAERI in waste treatment technology to cope with waste produced by pyroprocessing, which is used to effectively manage spent fuel. The goal of pyroprocessing waste treatment is to reduce final waste volume, fabricate durable waste forms suitable for disposal, and ensure safe packaging and storage. KAERI employs grouping of fission products recovered from process streams and immobilizes them in separate waste forms, resulting in product recycling and waste volume minimization. Novel aspects of KAERI approach include high temperature treatment of spent oxide fuel for the fabrication of feed materials for the oxide reduction process, and fission product concentration or separation from LiCl or LiCl-KCl salt streams for salt recycling and higher fission-product loading in the final waste form. Based on laboratory-scale tests, an engineering-scale process test is in progress to obtain information on the performance of scale-up processes at KAERI.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.217-225
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• 2020

Concrete waste accounts for approximately 70~80% of the total waste generated during the decommissioning of nuclear power plants (NPPs). Based upon the concentration of each radionuclide, the concrete waste from the decommissioning can be used in the determination of the clearance threshold used to classify waste as radioactive. To reduce the cost of radioactive concrete waste disposal, it is important to perform decontamination before self-disposal or limited recycling. Therefore, it is necessary to estimate the internal radioactivity distribution of radioactive concrete waste to ensure effective decontamination. In this study, the performance metrics of various Compton reconstruction algorithms were compared in order to identify the best strategy to estimate the internal radioactivity distribution in concrete waste during the decommissioning of NPPs. Four reconstruction algorithms, namely, simple back-projection, filtered back-projection, maximum likelihood expectation maximization (MLEM), and energy-deconvolution MLEM (E-MLEM) were used as Compton reconstruction algorithms. Subsequently, the results obtained by using these various reconstruction algorithms were compared with one another and evaluated, using quantitative evaluation methods. The MLEM and E-MLEM reconstruction algorithms exhibited the best performance in maintaining a high image resolution and signal-to-noise ratio (SNR), respectively. The results of this study demonstrate the feasibility of using Compton images in the estimation of the internal radioactive distribution of concrete during the decommissioning of NPPs.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.11a
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• pp.305-306
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• 2007

ACP(Advanced Spent Fuel Conditioning Process)의 금속전환로에 $U_3O_8$을 공급하기 위하여 20 kgHM/batch의 $UO_2$ 펠릿(pellets)을 처리할 수 있는 건식분말화 장치가 개발되고있다. 건식분말화 장치는 500 $^{\circ}C$온도에서 공기를 공급하여 일정한 입도범위의 균질한 $U_3O_8$을 만든다. 이런 건식 분말화 장치의 효율을 높이기 위해서는 반웅로에 불어 넣어주는 공기의 유량을 증가시킬 필요가 있다. 하지만 공기와 반응하여 생성되는 $U_3O_8$ 입자는 그 크기가 최소 3 ${\mu}$m 정도로 매우 미세하여，반응로 출구를 통해 외부로 빠져나갈 가능성 이있다. 이를 방지하기 위해 분말화 장치 출구 바깥에는 필터가 설치되어 있으나 공기와 함께 $U_3O_8$ 입자가 계속해서 빠져 나갈 경우 입자로 인해 필터가 막혀 제 기능을 할 수 없게 된다. 따라서 건식 분말화 장치는 미세한 $U_3O_8$ 입자가 반응로 밖으로 빠져나가지 않도록 입구에서의 공기 유량을 일정 수준 이하로 조절해주는 것이 필요하다. 이 연구의 목적은 초기 유량으로부터 유량을 점점 증가시키면서 시간변화에 따른 입자 거동 특성을 해석하며， 결과로부터 주어진 크기의 타원입자에 대해 최대 허용 공기 유량을 결정하고자한다. 이 해석을 위해 유동과 입자를 동시에 해석할 수 있는 ANSYS-CFX 5.7.1과 ANSYS-CFX 10.0 두 가지의 소프트웨어가 사용되었다. 해석 결과를 바탕으로 좀더 정확한 유량 한계치 계산을 위해 추가로 수행되어야 할 해석에 대해 제안하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.367-372
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• 1996

석탄화력발전소의 산업부산물인 fly ash를 이용한 폐흡착재의 붕규산유리고화가 능성을 분석하였다. 폐흡착재는 기체상의 세슘이나 루테늄 등을 포집한 후에 발생되는 필터류 등의 고체폐기물을 말하며 본 실험에서는 CsNO$_3$와 fly ash를 몰비로 1.5 : 1 되게 섞어 1200 $^{\circ}C$에서 1시간 가소 시킨 후에 생성되는 pollucite를 모의폐흡 착재로 사용하였다. 폐흡착재를 무게비 15 ~ 30 %로 fly ash, SiO$_2$, $Na_2$CO$_3$, B$_2$O$_3$와 혼합한 후 1150 $^{\circ}C$에서 3시간 용융시켜 붕규산유리화시켰다. 제조된 붕규산유리고화체의 침출성을 평가하기 위하여 2일동안의 soxhlet 침출실험을 수행하였다. 한편 폐흡착재의 붕규산유리고화과정을 알아보기 위하여 붕규산유리고화체의 원료물질에 대하여 유리화과정과 동일한 조건하에서 TG/DTA분석을 수행하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.303-309
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• 2010

For the pyrochemical process of spent nuclear fuels, recovery of LiCl-KCl eutectic salts is needed to reduce radioactive waste volume and to recycle resource materials. This paper is about recovery of residual LiCl-KCl eutectic salts in radioactive rare earth precipitates (rare earth oxychlorides or oxides) by using a vacuum distillation process. In the vacuum distillation test apparatus, the salts in the rare earth precipitates were vaporized and were separated effectively. The separated salts were deposited in three positions of the vacuum distillation test apparatus or were collected in the filter and it is difficult to recover them. To resolve the problem, a vacuum distillation and condensation system, which is subjected to the force of a temperature gradient at a reduced pressure, was developed. In a preliminary test of the vacuum distillation/condensation recovery system, it was confirmed that it was possible to condense the vaporized salts only in the salt collector and to recover the condensed salts from the salt collector easily.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.319-327
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• 2010

A nuclear plant ESF ACS simulator was designed, built, and verified to perform experiment related to ESF ACS of nuclear power plants. The dimension of 3D CAD model was based on drawings of the main control room(MCR) of Yonggwang units 5 and 6. The CFD analysis was performed based on the measurement of the actual flow rate of ESF ACS. The air flowing in ACS was assumed to have $30^{\circ}C$ and uniform flow. The flow rate across the HEPA filter was estimated to be 1.83 m/s based on the MCR ACS flow rate of 12,986 CFM and HEPA filter area of 9 filters having effective area of $610{\times}610mm^2$ each. When MCR ACS was modeled, air flow blocking filter frames were considered for better simulation of the real ACS. In CFD analysis, the air flow rate in the lower part of the active carbon adsorber was simulated separately at higher than 7 m/s to reflect the measured value of 8 m/s. Through the CFD analyses of the ACSes of fuel building emergency ventilation system, emergency core cooling system equipment room ventilation cleanup system, it was confirmed that all three EFS ACSes can be simulated by controlling the flow rate of the simulator. After the CFD analysis, the simulator was built in nuclear grade and its reliability was verified through air flow distribution tests before it was used in main tests. The verification result showed that distribution of the internal flow was uniform except near the filter frames when medium filter was installed. The simulator was used in the tests to confirm the revised contents in Reg. Guide 1.52 (Rev. 3).