The nuclear fuel that melted during the Fukushima nuclear accident in 2011 is still being cooled by water. In this process, contaminated water containing radioactive substances such as cesium and strontium is generated. The total amount of radioactive pollutants released by the natural environment due to the nuclear accident in Fukushima in 2011 is estimated to be 900 PBq, of which 10 to 37 PBq for cesium. Radioactive cesium (137Cs) is a potassium analog that exists in the water in the form of cations with similar daytime behavior and a small hydration radius and is recognized as a radioactive nuclide that has the greatest impact on the environment due to its long half-life (about 30 years), high solubility and diffusion coefficient, and gamma-ray emission. In this study, alginate beads were designed using Prussian blue, known as a material that selectively adsorbs cesium for removal and detection of cesium. To confirm the adsorption performance of the produced Prussian blue, immersion experiments were conducted using Cs standard solution, and MCNP simulations were performed by modeling 1L reservoir to conduct experiments using radioactive Cs in the future. An adsorption experiment was conducted with water containing standard cesium solution using alginate beads impregnated with Prussian blue. The adsorption experiment tested how much cesium of the same concentration was adsorbed over time. As a result, it was found that Prussian blue beads removed about 80% of cesium within 10-15 minutes. In addition, MCNP simulation was performed using a 1 L reservoir and a 3inch NaI detector to optimize the amount of Prussian blue. The results of comparing the efficiency according to the Prussian volume was shown. It showed that our designed system holds great promise for the cleanup and detection of radioactive cesium contaminated seawater around nuclear plants and/or after nuclear accidents. Thus, this work is expected to provide insights into the fundamental MCNP simulation based optimization of Prussian blue for cesium removal and this work based MCNP simulation will pave the way for various practical applications.

When decommissioning of nuclear facilities happens, large amounts of radioactive wastes are released. Because costs of nuclear decommissioning are enormous, effective and economical decontamination technologies are needed to remove radioactive wastes. During NPP operation, corrosion product called Chalk River Unidentified Deposits (CRUD) is generated. CRUD is an accumulation of substances and corrosion products consisting of dissolved ions or solid particles such as Ni, Fe, and Co on the surface of the NPP fuel rod coating. CRUD is slowly eroded by the circulation of hot pressurized water and later deposits on the fuel rod cladding or external housing, thereby reducing heat production by the nuclear fuel. Decontamination of radiologically contaminated metals must be performed before disposal, and several methods for decontaminating CRUD are being studied in many countries. Decontamination technology is an alternative to reducing human body covering and reducing radioactive waste disposal costs, and much research and development has been conducted to date. Currently, the importance of decontamination is emerging as the amount of waste stored in radioactive waste storage is close to saturation, and the amount of radioactive waste generated must be minimized through active decontamination. In this study, a preliminary study was conducted on the removal of CRUD by multiple membrane in an electro-kinetic process using an electrochemicalbased decontamination method. Preliminary research to develop a technology to electrochemically remove CRUD by using a self-produced electrochemical cell to check the pH change over time of the CRUD cell according to voltage, electrolyte, membrane and pH change.

The primary heat transport system consists mainly of the in-core fuel channels connected to the steam generators by a system of feeder pipes and headers. The feeders and headers are made of carbon steel. Feeders run vertically upwards from the fuel channels across the face of the reactor and horizontally over the refueling machine to the headers. Structural materials of the primary systems of nuclear power plants (NPPs) are exposed to high temperature and pressure conditions, so that the materials employed in these plants have to take into accounts a useful design life of at least 30 years. The corrosion products, mainly iron oxides, are generated from the carbon steel corrosion which is the main constituent of the feeder pipes and headers of this circuit. Typical film thickness on CANDU-PHWR surface is 75μm or 30mg/cm2. Deposits on PHWR tends to be much thicker than PWR due to use of carbon steel and also for the source of corrosion products available on the carbon steel surface. Degradation of carbon steel for the feeder pipes transferring the primary system coolant by flow-assisted corrosion in high temperature has been reported in CANDU reactors including Point Lapreau, Gentully-2, Darlington and Bruce NPPs. The formation of Fe3O4 film on a carbon steel surface reduces the dissolution rate of steel substantially. The protectiveness of the Fe3O4 film over the carbon steel is affected by the environmental factors and the operational parameters of the feeder pipes, including the velocity, wall shear stress, solution pH, temperature, concentration of dissolved iron, quality of solution, etc. For effective chemical decontamination of these thick oxides containing radionuclides such as Co-60, it is necessary to understand the corrosion behaviors of feeder pipes and the characteristics of oxide formed on it. In this work, we investigated the growth of oxide films that develop on type SA-107 Gr. B carbon steel in high temperature water and steam environment by scanning electron microscopy (SEM) and glow discharge optical emission spectrometry (GD-OES) for the quantification and the solidstate speciation of metal oxide films. This study was especially focused to set the experimental tests conditions how to increase the oxide thickness up to 50 m by changing the oxidation conditions, such as solution chemistry and thermo-hydraulic conditions both temperature and pressure and so on.

It is crucial to be sure about the safety of nuclear facilities for human resources who are in danger of radioactive emission, also diminishing the volume of the wastes that are buried under the ground. Chemical decontamination of nuclear facilities can provide these demands at the same time by dissolving the oxide layer, which radionuclides such as 60Co and 58Co have been penetrated, of parts that are utilized in nuclear plants. Although there are many commercial methods to approaching its aim and they perform a high decontamination factor, they have some issues such as applying organic acids which have the ability to chelate with radionuclides that can be washed by underground water, have large quantities of radioactive waste and damage to the surface by severe intergranular attack. A new method has been introduced by KAERI’s scientist which is named the HyBRID Process, in this process the main solution is the acidic form of Hydrazine. In this process, like other acid-washing processes, there is a chance of corrosion on the metal surface which is not desired. The metal surface is able to be protected during dissolving process by adding some organic and inorganic corrosion inhibitors such as PP2 and PP3. There is a very new research topic about ionic liquids (ILs) as corrosion inhibitors which illustrates a vast potential for this application due to their tunable nature and the variety of options for cationic and anionic parts. The key factors for ILs corrosion inhibitors such as the hardness properties are summarized. In this study, we review to the fundamentals and development of corrosion inhibitors for chemical decontamination and give an prospect with emphasis on the challenges to be overcome.

The radiation field generated in the primary cooling system of a nuclear power plant tends to increase in intensity as radionuclides bind to the oxide film on the internal surface of the primary system, which is operated at high temperature and pressure, and as the number of years of operation increases. Therefore, decontamination of the primary cooling system to reduce worker exposure and prevent the spread of contamination during maintenance and decommissioning of nuclear power plants uses the principle of simultaneous elution of radionuclides when the corrosion oxide film dissolves. In general, a multi-stage chemical decontamination process is applied, taking into account the spinel structure of the corrosion oxide film formed on the surface of the primary cooling system, i.e. an oxidative decontamination step is applied first, followed by a reductive decontamination step, which is repeated several times to reach the desired decontamination goal. Currently, permanganic acid is commonly used in oxidative decontamination processes to remove Cr from corrosion oxide films. In the reductive decontamination step to remove iron and nickel, organic acids such as oxalic acid are commonly used. However, organic acids are not suitable for the final radioactive waste form. A number of multi-stage chemical decontamination technologies for primary cooling systems have been developed and commercialized, including NP-CITROX, AP/NP-CANDECON, CANDERM, AP/NP-LOMI and HP/CORD-UV. Among these, HP/CORDUV is currently the most actively applied primary cooling system chemical desalination process in the world. In this study, KAERI has developed a new chemical decontamination technology that does not contain organic chemical decontamination agents, with a focus on securing an original technology for reducing the amount of decontamination waste while having equivalent or better decontamination performance than overseas commercial technologies, and compared it with the inorganic chemical agent-based HyBRID (Hydrazine Based Reductive Metal Ion Decontamination) chemical decontamination technology.

There are analytical methods used for measuring activity when light photons are emitted for scintillating-based analytical application. When this electron returns to the original stable state, it releases its energy in the form of light emission (visible light or ultraviolet light), and this phenomenon is called scintillation. Scintillator is a general term for substances that emit fluorescence when exposed to radiation such as gamma-rays. Radioactivity is all around us and is unavoidable because of the ubiquitous existence of background radiations emitted by different sources. The scintillator contributes to these sensing, and it is expected that the inspection accuracy and limit of detection will be improved and new inspection methods will be developed in the future. Moreover, scintillators are chemical or nanomaterial sensors that can be used to detect the presence of chemical species and elements or monitor physical parameters on the nanoscale. In this study, it includes finding use in scintillating-based analytical sensing applications. A chemical and nanomaterial based sensors are self-contained analytical tools that could provide information about the chemical compositions or elements of their environment, that is, a liquid or even gas condition. Herein, we present an insightful review of previously reported research in the development of high-performance gamma scintillators. The major performance-limiting factors of scintillation are summed up here. Moreover, the 2D material has been discussed in the context of these parameters. It will help in designing a prototype nanomaterial based scintillators for radiation detection of gamma-ray.

Radioactive contamination distribution in nuclear facilities is typically measured and analyzed using radiation sensors. Since generally used detection sensors have relatively high efficiency, it is difficult to apply them to a high radiation field. Therefore, shielding/collimators and small size detectors are typically used. Nevertheless, problems of pulse accumulation and dead time still remain. This can cause measurement errors and distort the energy spectrum. In this study, this problem was confirmed through experiments, and signal pile-up and dead time correction studies were performed. A detection system combining a GAGG sensor and SiPM with a size of 10 mm × 10 mm × 10 mm was used, and GAGG radiation characteristics were evaluated for each radiation dose (0.001~57 mSv/h). As a result, efficiency increased as the dose increased, but the energy spectrum tended to shift to the left. At a radiation dose intensity of 400 Ci (14.8 TBq), a collimator was additionally installed, but efficiency decreased and the spectrum was distorted. It was analyzed that signal loss occurred when more than 1 million particles were incident on the detector. In this high-radioactivity area, quantitative analysis is likely to be difficult due to spectral distortion, and this needs to be supplemented through a correction algorithm. In recent research cases, the development of correction algorithms using MCNP and AI is being actively carried out around the world, and more than 98% of the signals have been corrected and the spectrum has been restored. Nevertheless, the artificial intelligence (AI) results were based on only 2-3 overlapping pulse data and did not consider the effect of noise, so they did not solve realistic problems. Additional research is needed. In the future, we plan to conduct signal correction research using ≈10×10 mm small size detectors (GAGG, CZT etc.). Also, the performance evaluation of the measurement/analysis system is intended to be performed in an environment similar to the high radiation field of an actual nuclear facility.

우라늄 토양 및 콘크리트 폐기물의 동전기 제염 후 방사성폐기물의 시멘트 고화특성을 분석하기 위하여, 시멘트 고화 유동성 시험을 수행하고 시멘트 고화 시료를 제작하였다. 시멘트 고화시료에 대하여 압축강도, pH, 전기전도도, 방사선조사 효과 및 부피증가를 분석하였다. 방사성폐기물의 시멘트 고화의 작업 적정도는 175~190% 정도였다. 시멘트 고화시료의 방사선 조사 후 압축강도는 방사선 조사 전 압축강도 보다 약 15% 감소하였으나, 한국원자력환경공단 인수기준 (34 kgf·cm-2)을 만족하였다. 동전기 제염 후 방사성폐기물의 시멘트 고화 시료에 대한 SEM-EDS 분석결과, 알루미늄상은 시멘트와 잘 결합 한 형상을 나타낸 반면, 칼슘상은 시멘트와 분리된 형상을 나타내었다. 방사성폐기물의 시멘트 고화 부피는 시멘트에 대한 폐기물의 배합과 수분량에 따라 다르게 나타났다. 방사성폐기물의 시멘트 고화 부피(C-2.0-60)는 약 30% 증가였으며 동전기 제염 후 생성된 방사성폐기물의 영구처분은 적절하다고 판단되었다.

Nuclide separation of commercial nanofiltration (NF) & reverse osmosis (RO) membranes is investigated to elucidate the permeation mechanism and possibility on applying treatment of wastewater from nuclear facility. Since wastewater from nuclear facility contains significant amount of dangerous nuclides, their treatment for recycle or clearance is necessary. Polyamide based commercial membranes are prepared to investigate their separation performance on single- & multi-component nuclide aqueous solutions. Similar to their desalination performance, NF shows higher flux, but RO has high rejection. With respect to rejection, NF and RO membranes have better performance comparing to desalination. This study indicates that NF and RO membranes can be potential candidate for nuclide separation.