탄소섬유 강화 플라스틱 (Carbon fiber reinforced plastics, CFRP)은 고함량의 탄소섬유 (Carbon fiber, CF)와 고분자로 이루어진 복합재료로서, 뛰어난 기계적 성능으로 항공우주, 자동차, 토목 등 다 양한 산업 분야에서 사용되고 있다. 하지만 사용량 증가에 따른 폐기물의 환경문제와 추출한 재활용 탄소섬유 (Recycled carbon fiber, rCF)의 적용 가능 분야의 한계로 인해 재활용이 제한적인 실정이 다. 본 연구에서는 rCF와 CF 혼입 시멘트계 전자파 복합재를 제작하여 그 성능을 비교 분석하기 위 한 실험을 수행하였다. 구성재료는 시멘트, 잔골재, 고성능 감수제를 사용하였으며, 비교 분석을 위해 CF와 rCF를 각각 6 mm, 12 mm 길이를 0.1, 0.3, 0.5, 1.0 wt.% 함량으로 사용하였다. 전자파 복합 재의 흡수 성능 향상을 위해 각각 다른 함량의 다층 구조를 형성하였으며, 전자파 투과를 낮은 함량에 서 높은 함량 방향이 되도록 측정을 진행하였다. 전자파 차폐성능은 재령 28일 이후 네트워크 분석기 를 사용하여 자유 공간에서 측정하였으며, C-band (4~8 GHz)와 X-band (8~12 GHz) 주파수 영역 에서의 반사율과 투과율을 각각 측정하였다.

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO2 fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

Radiation dose rates for spent fuel storage casks and storage facilities of them are typically calculated using Monte Carlo calculation codes. In particular, Monte Carlo computer code has the advantage of being able to analyze radiation transport very similar to the actual situation and accurately simulate complex structures. However, to evaluate the radiation dose rate for models such as ISFSI (Independent Spent Fuel Storage Installation) with a lot of spent fuel storage casks using Monte Carlo computational techniques has a disadvantage that it takes considerable computational time. This is because the radiation dose rate from the cask located at the outermost part of the storage facility to hundreds of meters must be calculated. In addition, if a building is considered in addition to many storage casks, more analysis time is required. Therefore, it is necessary to improve the efficiency of the computational techniques in order to evaluate the radiation dose rate for the ISFSI using Monte Carlo computational codes. The radiation dose rate evaluation of storage facilities using evaluation techniques for improving calculation efficiency is performed in the following steps. (1) simplified change in detailed analysis model for single storage cask, (2) create source term for the outermost side and top surface of the storage cask, (3) full modeling for storage facilities using casks with surface sources, (4) evaluation of radiation dose rate by distance corresponding to the dose rate limit. Using this calculation method, the dose rate according to the distance was evaluated by assuming that the concrete storage cask (KORAD21C) and the horizontal storage module (NUHOMS-HSM) were stored in the storage facility. As a result of calculation, the distance to boundary of the radiation control area and restricted area of the storage facility is respectively 75 m / 530 m (KORAD21C case), and 20 m / 350 m (NUHOMS-HSM case).

The Co-60 is a radioactive material widely used in domestic and foreign medical, industrial, health and research fields. Currently, world market for the Co-60 is about 80 MCi/yr and is expected to grow to 150 MCi/yr by 2025. For the Co-60, Nordion of Canada occupies about 80% of the world market. In the case of Korea, a small amount of sources with relatively low radioactivity intensity are produced using research reactors, but most of the Co-60 is entirely dependent on imports. Accordingly, although the technical feasibility of the Co-60 production technology using the PHWR was evaluated, it was evaluated as a negative result on the additional construction of a hot cell, core management, safety analysis and economic feasibility. Canada, the main producer of the Co-60, is also conducting research on the Co-60 production technology using PWR with GE-Hitachi and Westinghouse as the number of PHWR is expected to decrease. In Korea, it is necessary to preoccupy the Co-60 production technology and auxiliary technology using the PWR by utilizing excellent technology, and active research is being conducted to secure unique nuclear power technology that does not depend on foreign countries. Therefore, in this study, the thickness and weight of the radioactive shielding required for handling (transport) of Co-60 produced using the PWR were calculated.

The spent filters stored in Kori Unit 1 are planned that compressed and disposed for volume reduction. However, shielding reinforcement is required to package high-dose spent filters in a 200 L drum. So, in this study suggests a shielding thickness that can satisfy the surface dose criteria of 10 mSv·h−1 when packaging several compressed spent filters into 200 L drums, and the number of drums required for the compressed spent filter packaging was calculated. In this study, representative gamma-emitting nuclides in spent filter are assumed that Co-60 and Cs-137, and dose reduction due to half-life is not considered, because the date of occurrence and nuclide information of the stored spent filter are not accurate. The shielding material is assumed to be concrete, and the thickness of the shielding is assumed to 18 cm considering the diameter of the spent filter and compression mold. Considering the height of the compressed spent filter and the internal height of the shielding drum, assuming the placement of the compressed spent filter in the drum in the vertical direction only, the maximum number of packaging of the compressed spent filter is 3. When applying a 18 cm thick concrete shield, the maximum dose of the spent filter can packaged in the drum is 125 mSv·h−1, so when packaging 3 spent filters of the same dose, the dose of a spent filter shall not exceed 41 mSv·h−1 and not exceed 62 mSv·h−1when packing 2 spent filters. Therefore, the dose ranges of spent filters that can be packaged in a drum are classified into three groups: 0–41 mSv·h−1, 41–62 mSv·h−1, and 62–125 mSv·h−1based on 41 mSv·h−1, 62 mSv·h−1, and 125 mSv·h−1. When 227 spent filters stored in the filter room are classified according to the above dose group, 207, 3 and 4 spent filters are distributed in each group, and the number of shielding drums required to pack the appropriate number of spent filters in each dose group is 75. Meanwhile, 8 spent filters exceeding 125 mSv·h−1 and 5 spent filters that has without dose information are excluded from compression and packaging until the treatment and disposal method are prepared. In the future, we will segmentation of waste filter dose groups through the consideration of dose reduction and horizontal placement of compressed spent filters, and derive the minimum number of drums required for compressed spent filter packaging.

In this study, we quantitatively analyzed the data by measuring the radiation shielding rate and uniformity in order to evaluate the performance of an Apron. In addition, storage conditions were also evaluated. The uniformity measurement was performed by evaluating the Apron DICOM images using a PACS program. The experiment was intended for 51 Aprons being used in three hospitals in the Daejeon area. The radiation shielding rate and uniformity were measured per lead equivalent for 0.25 mmPb, 0.35 mmPb, and 0.5 mmPb. As a result, the higher lead equivalents were, the greater differences in the non-uniformity between the top part and the bottom part became (p=0.020). In all hospitals, regarding the non-uniformity of four places in Aprons, all showed statistically significant differences (p<0.01). The average value of the transmitted radiation dose showed less difference (p = 0.005) in the bottom right than in the upper right but was statistically significant. There have been no marks of manufacturing date or the date of purchase in the Apron.

한국원자력연구원 내 저장중인 중·저준위방사성폐기물의 안전한 영구처분을 위하여 방사성폐기물처 리시설에서는 이들 저장폐기물의 핵종재고량 평가를 위한 드럼핵종분석장치의 도입을 기획하고 있으며, 이 장치를 운영하고 유지보수하기 위한 전용의 평가시설 구축을 추진하고 있다. 이 평가시설 외부에는 제 1방사성폐기물저장시설이 인접하여 있으며 내부에는 평가대상드럼들과 드럼핵종분석장치의 밀도보정용 선원 등이 존재함으로 이들이 방사선원항으로 작용하여 평가시설 내·외부의 선량률에 영향을 미치게 된 다. 따라서 방사선원항 주변의 콘크리트 구조물에 대한 방사선차폐 영향을 평가하여야 한다. 본 연구에서 는 MCNP 코드를 이용하여 해당 방사선원항으로부터 평가시설 내부 방사선관리구역을 둘러싸고 있는 콘 크리트 구조물에 대한 방사선 안전성을 평가하였다. 평가결과 현재 고려되고 있는 콘크리트 구조물의 약 30 cm 두께는 해당 방사선원항으로부터의 방사선을 차폐하기에 충분한 것으로 확인되었다.

한국원자력연구소에서는 고온의 용융염 매질 하에서 사용 후 핵연료를 환원시키는 차세대관리종합공정 연구를 수행 중에 있다. 추후 본 기술개발을 실증시험 하기 위해서는 방사선 차폐능이 확보된 핫셀이 필수적이며, 핫셀은 최대 1,385TBq의 방사능량에 대한 차폐 안전성을 가져야 한다. 최대 방사선원에 대한 핫셀의 차폐능을 확보하기 위하여, 본 연구에서는 실증시험 시 사용후핵연료부터 발생하는 중성자 및 감마선에 의한 선량률이 법적 허용선량치보다 낮게 유지되도록 핫셀의 차폐 설계에 대한 안전성을 평가하였다. QAD-CGGP 및 MCNP-4C 코드를 이용하여 핫셀 차폐체의 설계치에 대한 차폐 계산을 수행하였다. 작업구역에 대한 감마선 차폐계산 결과 QAD-CGGP 코드는 2.10, 2.97 mSv/h, MCNP-4C 코드는 1.60, 2.99 mSv/h 이었으며, 서비스 구역은 1.01, 7.88 mSv/h 로 평가되었다. 그리고 MCNP-4C코드를 이용하여 중성자에 의한 선량률을 계산한 결과, 중성자에 의한 선량률은 감마에 의한 선량률의 약 20% 이하치를 나타내었다. 따라서 선량률 대부분은 감마선에 의한 영향임을 알 수 있었다. 본 연구를 통하여 핫셀의 차폐 설계치가 작업구역의 선량 제한치 0.01 mSv/h 와 서비스 구역에서의 선량 제한치 0.15 mSv/h를 만족시키는 것을 확인할 수 있었다.

금속을 함유하고 있는 산업폐자원과 관련한 재활용 기술을 개발하기 위해 다양한 접근이 시도되고 있으며, 그 중에서 유리는 미생물로 분해되지 않기 때문에 매립은 적합하지 않아 폐유리의 재활용에 대한 관심은 증대되고 있다. 따라서 본 논문에서는 폐유리를 잔골재로 사용하고 폐유리의 중금속 용출을 억제하기 위한 킬레이트 수지를 혼입함으로써, 차폐 채움재의 강도, 건조수축, 알칼리-실리카반응, 중금속 용출 등을 평가하여 폐유리를 경제적이며 환경 친화적인 차폐 채움재로서 활용하기 위한 기초자료를 제시하고자 한다. 시험결과, 폐유리를 잔골재로 사용하였을 경우 강도 발현에 효과적이었으며, 킬레이트 수지를 혼입하였을 경우 강도 발현에 영향이 있는 것으로 나타났다. 또한 킬레이트 수지를 혼입하였을 경우 건조수축의 개선에는 효과적이었으나 알칼리-실리카반응에 영향을 미치는 것으로 나타났다. 중금속 용출 시험결과, 한국 KSLP 시험법에서는 중금속 용출 허용 기준치를 모두 만족하였으나, 납의 경우 미국 ANSI 67-2007a의 허용 기준치를 초과하여 이에 대한 추가적인 연구가 진행되어야 할 것으로 판단되었다.