The operation and decommissioning of nuclear power plants (NPPs) creates waste in the process of handling radioactively contaminated material, which must be disposed of in a repository or can be recovered of in the same way as conventional waste in the course of handling radioactively contaminated materials. For buildings or sites of NPPs it also has to be decided under what conditions they can continue to be used for other, conventional purposes or demolished. This decision is referred to as “release from supervision under nuclear and radiation protection law” or “clearance” in short. The clearance levels applicable in Germany according to the Radiation Protection Ordinance have been defined such that a radiation dose (hereinafter referred to as “dose”) of 10 μSv per year is not exceeded. The vast majority of the materials resulting from the dismantling of a nuclear power plant (e.g. most of the massive concrete structures) are neither contaminated nor activated. Thus, there is no need to treat these materials as radioactive waste. Emplacement of uncontaminated masses which in Germany is essentially several million tonnes of building rubble in a repository would require additional construction of such facilities, which, in view of the negligible hazard potential, from the point of view of the Nuclear Waste Management Commission (ESK) is clearly to be rejected both economically and, in particular, ecologically. Alternative ways are increasingly discussed in public, such as the abandonment of buildings after gutting, i.e. refraining from demolition of the controlled area buildings of NPPs. Also, another proposal discussed in public, the landfilling or the long-term storage of cleared material at the site, does not offer any safety-related advantages either in the view of the ESK. If, after completion of all dismantling work, the building has been decontaminated such that the clearance levels for buildings are complied with further use of the building rubble resulting from demolition is harmless from a radiological point of view. For these reasons, Germany has deliberately decided to use clearance as an essential measure in the dismantling of NPPs. If it is intended to conventionally reuse or depose of virtually contaminant-free material from controlled areas, it must first undergo a clearance procedure. The prerequisites that must be fulfilled for clearance are regulated in the Radiation Protection Ordinance, which includes two basic clearance pathways: unrestricted and specific clearance. In the following, the basic process of clearance is briefly presented and illustrated for a better understanding. It comprises five steps. Step 1-Radiological characterization by sampling, Step 2-Dismantling of plant components in the controlled area, Step 3- Decontamination, Step 4-Decission measurements, Step 5-Clearacnce and further management. The entire clearance process is governed by a clearance notice and is carried out under the supervision of the competent authority under nuclear and radiation protection law or the independent authorized expert commissioned by it. The clearance pathways contained in the Radiation Protection Ordinance have proven themselves in practice. They permit safe and proper management of material from dismantling and release of the site from supervision under nuclear and radiation protection law. These German regulatory procedures should be taken into account and deregulation and removal should be used as appropriate and necessary tools in the process of decommissioning NPPs in order to return non-hazardous materials to the material cycle or for conventional disposal.

Considering the Fukushima nuclear accident and the marine discharge plan of contaminated (or treated) water, it is necessary a seafood monitoring system for radioactive material screening. Currently, radioactivity tests in seafood are conducting in Korea. Although current method using a HPGe detector can provide very low uncertainty in determining radioactivity, there is a limitation in that rapid inspection cannot be performed because of a time-consuming pretreatment process as well as long measurement time (typically 10,000 s). To overcome this limitation, we are developing an insitu inspection device, a kind of screening system, which can monitor the radioactivity in seafood in a near real-time basis. In this study, the actual seafood with a check source was measured to verify the reliability of the Monte Carlo simulation model. The detector used in the experiment was a 5-cm-thick polyvinyl toluene (PVT) plastic scintillator with a 0.5-cm-thick lead shield for background reduction. A Cs-137 check source was placed within seafood. The seafood used in the experiment was fishcake, raw oyster, and dried laver, which is the representative seafood of fish, shellfish, and seaweed. These three seafood products of the same size and shape as the manufacturing process were used to predict the performance realistically. We compared the energy spectrum of the Cs-137 check source obtained from measurements and simulations. The region of interest (ROI) around the Compton edge was set to reduce the influence of the background and electronic noise. The results showed that the measured and simulated spectrum were in good agreement.

Radioactive mixed waste (RMW) is containing radioactive materials and hazardous materials. Radioactive wastes containing asbestos are include in RMW. These wastes thus must be treated considering both radioactive and hazardous aspects. In this study, a high temperature melt oxidation system consisting of an electric arc furnace and a molten salt oxidation furnace has been developed for the treatment of of radioactive waste containing asbestos. A surrogate waste of the radioactive waste containing asbestos (content of asbestos: 13wt%) was treated in this system. It was melted and fabricated into a glass waste form in the system. Asbestos was not detected in this glass waste form. This means that the asbestos was converted to a glass component in the glass waste form. The waste form was homogeneous glass, and it had a high value of compressive strength (475.13 MPa). It was also confirmed through a leaching test (ANS 16.1) that the waste form had a high chemical durability (Leaching Index > 6). Based on these results, it is considered that the high temperature melt oxidation system will be utilized for the treatment of a significant amount of radioactive waste containing asbestos generated from decommissioning a nuclear power plant.

The dismantlement of the Kori Unit 1 and Wolsong Unit 1 nuclear power plants is scheduled. Since about 40% of the cost of dismantling nuclear power plants is the cost of disposing of generated wastes, it is important to secure recycling technologies. Among them, low and intermediate level radioactive wastes are made of porous filters and adsorbent materials of ceramic foam to remove nuclides such as C-14, I, and Xe generated during nuclear dismantling. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. Therefore, in this study, the foam filter performance was improved by applying a sacrificial material to increase the specific surface area and porosity of the ceramic foam filter. The sacrificial material is burned out with polyurethane (PU) of the green filter before the heat treatment temperature to increase the strength of the ceramic foam filter so that it can be maintained as pores, thereby improving the specific surface area and porosity. The sacrificial materials and melting temperature (Tm) reviewed in this study were anthracite (530~660°C), PMMA (160°C), Cellulose acetate (260~270°C), and aluminum particle (660°C), and their effect on the manufacture of foam filters was studied by applying this. The specific surface part and porosity of the foam filter were improved when anthracite and aluminum particle were added, and PMMA and Cellulose acetate, which are relatively low temperature melting points, were burned out at a temperature lower than PU, and thus their physical properties were not greatly affected. The physical properties and specific surface part and porosity of ceramic foam filters manufactured using various sacrificial materials will be discussed.

Barrier effect model developed by CRIEPI is used for the estimation of rate of radioactive material release from a transport cask submerged in the ocean. If the containment boundary of cask is broken in an accident during maritime transportation, the sea water comes into the cask cavity and the leaching of radioactive material occurs. If the release of radioactive material thorough the opening of the containment boundary of cask is less than the leaching rate of the radioactive material inside the cask, then the release rate is controlled by the saturation limit of the sea water inside the cask cavity. In this study, the release rate estimation using the barrier effect model is compared with the model used in other codes, such as MARINRAD. And by parameter study, important factors that affect the releaser rate are identified and prioritized. It is shown that the gap generated in the containment boundary is the key parameter that determine the release rate of the radioactive material and the leaching rate is the dominant parameter to determine the saturation time of the cavity sea water.

In accordance with the Enforcement Decree of the Act on Physical Protection and Radiological Emergency, operators of Nuclear Power Plants (NPP)s must conduct full cyber security exercise once a year and partial exercise at least once every half year. Nuclear operators need to conduct exercise on systems with high attack attractiveness in order to respond to the unauthorized removal of nuclear or other radioactive material and sabotage of nuclear facilities. Nuclear facilities identify digital assets that perform SSEP (Safety, Security, and Emergency Preparedness) functions as CDA (Critical Digital Assets), and nuclear operators select exercise target systems from the CDA list and perform the exercise. However, digital assets that have an indirect impact (providing access, support, and protection) from cyber attacks are also identified as CDAs, and these CDAs are relatively less attractive to attack. Therefore, guidelines are needed to select the exercise target system in the case of unauthorized removal of nuclear or other radioactive material and sabotage response exercise. In the case of unauthorized removal of nuclear or other radioactive material, these situations cannot occur with cyber attacks and external factors such as terrorists must be taken into consideration. Therefore, it is necessary to identify the list of CDAs that terrorists can use for cyber attacks among CDAs located in the path of stealing and transporting nuclear material and conduct intensive exercise on these CDAs. A typical example is a security system that can delay detection when terrorists attack facilities. In the case of sabotage exercise, a safety-related system that causes an initiating event by a cyber attack or failure to mitigate an accident in a DBA (Design Basis Accident) situation should be selected as an exercise target. It is difficult for sabotage to occur through a single cyber attack because a nuclear facility has several safety concepts such as redundancy, diversity. Therefore, it can be considered to select an exercise target system under the premise of not only a cyber attack but also a physical attack. In the case of NPPs, it is assumed that LOOP (Loss of Offsite Power) has occurred, and CDA relationships to accident mitigation can be selected as an exercise target. Through exercise on the CDA, which is more associated with unauthorized removal of nuclear or other radioactive material and sabotage of nuclear facilities, it is expected to review the continuity plan and check systematic response capabilities in emergencies caused by cyber attacks.

In this paper, as the transport cask was moved in the reactor, the structural integrity on the cask had to be evaluated in the normal transport condition. The drop height of the cask was determined by the weight of the cask in the normal transport condition by regulations about assessment test. It was determined that the drop height of the cask was 1.2 m by regulations. The velocity of the drop impact was calculated to perform the drop impact analysis by the principle of the conservation of energy. Using results of the simulation about the drop impact analysis, the structural integrity assessment on the transport cask was performed by ASME Boiler and Pressure Vessel Code.

IAEA 및 국내의 방사성물질 운반 관련 규정에 따라 중·저준위 방사성폐기물 드럼 8개를 운반할 수 있는 IP-2형 운반용기를 개발하였다. IP-2형 운반용기는 낙하시험 및 적층시험을 거친 후 내용물의 유실 또 는 분산과 운반용기 외부표면에서의 방사선량률이 20 % 이상 증가할 수 있는 차폐능력의 상실이 없어야 한다. 본 연구의 목적은 적층시험조건에 대한 시험방법 및 절차를 수립하고 IP-2형 운반용기의 적층조건 에 대한 구조적 건전성을 평가하는데 있다. 운반용기의 원형시험모델을 이용하여 운반용기 중량의 5배 하중으로 24시간 동안 압축하는 적층조건에 대한 시험 및 전산해석을 수행하였다. 적층시험 시 운반용기 의 모서리기둥에서의 변형률 및 변위를 측정하였으며, 측정된 변형률 및 변위는 해석결과와 서로 일치하 였다. 컨테이너 바닥부의 처짐량은 측정이 어렵기 때문에 전산해석 방법으로 구하였다. 모서리기둥의 최 대 변위와 컨테이너 바닥의 최대 처짐은 법규에서 규정하는 허용치에 비하여 낮게 나타났다. 적층시험 전?후에는 운반용기의 외형치수, 차폐체 두께, 볼트토크 등을 측정하였으며, 그 값들을 비교분석한 결과 운반용기는 내용물의 유실 및 분산, 차폐체 두께의 감소가 나타나지 않았다. 따라서 적층시험조건에서 IP-2형 운반용기의 구조적 건전성이 입증되었다.

본 연구는 사용 후 핵연료의 금속전환 공정에서 발생되는 폐용융염을 고형화하는 방법으로 실리카 함유 무기물을 이용하여 폐용융염을 열적, 수화학적 안정한 화합물로 전환하는 방법을 제안하였다. 실리카 함유 무기물(SAP)은 일반적인 sol-gel process로 합성되었으며, 및 로 구성된다. 제조된 SAP을 에서 폐용융염과 반응시켜 각 금속염화물에 대한 반응특성 및 열안정성을 조사하고, PCT 침출시험법을 이용하여 수화학적 안정성을 평가하였다. LiCl은 와 로, CsCl는 CS-aluminosilicate와 로, 는 로, 는 로 전환되었다. 9시간 동안 반응시킨 후, 금속염화물의 전환율은 였으며, 까지 열감량은 1wt%이하로 TGA(Thermo Gravimetric Analysis)로 확인하였다. Cs 및 Sr의 침출속도는 로 매우 높은 내침출특성을 나타내었다. 이상의 결과로부터, SAP으로 명명된 안정화제(stabilizer)는 금속염화물로 구성된 폐용융염에 대해 매우 효과적인 것으로 판단된다. SAP을 이용한 폐용융염의 고화처리방법은 후속적인 안정성의 검증과정을 통하여 폐용융염의 최종처분부피를 최소화할 수 있는 대안적인 고화방법으로 고려될 수 있을 것으로 기대 된다.