• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.426-427
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• 2005

Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical process in which the spent fuel of PWR is transformed into the uranic metal ingot. Through this process, which has been developed in KAERI since 1998, the radioactivity, the radiotoxicity, the heat and the volume of the PWR spent fuel are reduced by a quarter of the original. To demonstrate a lab-scale process and extract the data for the later pilot-scale process, a demonstration facility of ACP (ACPF) is under construction and the lab-scale demonstration is slated for 2006. To establish the safeguardability of ACPF, a safeguards system including a neutron counter based on non-destructive assay, which is named as ACP Safeguards Neutron Counter (ASNC), the ACP Safeguards Surveillance System (ASSS) which consists of two neutron monitors and five IAEA cameras, and Laser Induced Breakdown System (LIBS) have been developed and are ready to be installed at ACPF. The target materials of ACP to assay with ASNC are categorized into three types among which the first is the uranic metal ingot, the second is the salt waste and the last is $UO_2$ and $U_{3}O_8$ powders, rod cuts and hulls. The Pu content of process nuclear materials can be accounted with ASNC. The ASSS is integrated in the ACP Intelligent Surveillance Software (AISS) in which the IAEA camera images and background signals at the rear doors of ACPF are displayed. The composition of special nuclear materials of ACP can be measured with LIBS which can be a supporting measurement tool for ASNC. The conceptual picture of safeguards system of ACPF is shown in Fig. 1.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.91-97
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• 2010

As nuclear power plants are getting older, interests on a decontaminating process are increasingly attracting more attention. Chemical decontamination is crucial to lower the production of radioactive waste and radiation dose rate. Prior to this, oxidizers and detergents for target material should be chosen so as to decontaminate major systems and components of a nuclear power plant chemically. In order to decontaminate it properly, it is crucial to have information about the chemical composition and crystalline structure of CRUD, analyzing its samples from the target or the decontamination system with components. However, there is no program which enables the extraction of samples directly from the object or the decontamination system with components carrying genuine radioactivity. Therefore, it is limited to samples from corrosion products carrying partial radioactivity as a resource. The composition of CRUD varies considerably depending on refueling cycle because it is closely related to the constituent of basic material. After settling a target, it is crucial to analyze and obtain analytical information about CRUD as a decontamination target. In this paper, various technologies for manufacturing simulated CRUD are introduced as alternatives to unattained samples. A metal oxide or metal hydroxide was used to synthesize simulated cruds having chemical compositions and crystalline stricture similar to the actual one by 12 different methods. CRUD 4(metal oxides in the autoclave vessel) and CRUD 10(metal oxides in a crucible after hydrazing pretreatment)were chosen as the best method for Type 1 and Type 2.respectively. As these CRUD can be synthesized easily without using any specialized equipment or reagents in a short time and in large quantities, they are expected to stimulate the development of decontaminating agents and processes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.3
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• pp.187-199
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• 2015

Decontamination is one of the crucial technologies that are applied during the decommissioning process of nuclear facilities to secure the safety of workers and to minimize the quantity of radioactive waste. Decontamination removes radionuclides on the surface of contaminated metal. Compared with decontamination for operational nuclear facilities, decontamination for nuclear power plants that are being decommissioned needs to remove the more and thicker surface using more aggressive agents or specially developed equipment. This paper analyzed the factors to be considered before planning the decontamination, representative decontamination technologies, and their application procedure,etc. ORCID

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.149-155
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• 2005

The $ACP^1$ is under development for effective management of spent fuel by converting $UO_2$ into U-metal. For demonstration of this process, $\alpha-\gamma$ type new hotcell was built in the $IMEF^2$ basement. To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hotcell and service area at back of it. This system consists of 7 parts; Area Monitor for $\gamma$-ray, Room Air Monitor for particulate and iodine in both area, Hotcell Monitor for hotcell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.395-400
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• 2003

For the long term storage safety study of the metallic spent fuel, U-Nb, U-Ti, U-Ni, U-Zr, and U-Hf simulated metallic uranium alloys, known as corrosion resistant alloys, were fabricated and oxidized in oxygen gas at $200^{\circ}C~300^{\circ}C$. Simulated metallic uranium alloys were more corrosion resistant than pure uranium metal, and corrosion resistance increases Nb, Ni, Ti in that order. The oxidation rates of uranium alloys determined and activation energy was calculated for each alloy. The matrix microstructure of the test specimens were analyzed using OM, SEM, and EPMA. It was concluded that Nb was the best acceptable alloying elements for reducing corrosion of uranium meta] considered to suitable as candidate.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.421-425
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• 2003

The reductive extraction process is an important step to refine the TRU product from the electrorefining process for the preparation of transmutation reactor fuel. In this study, it was studied on the reductive extraction between the eutectic salt and Bi metal phases. The solutes were zirconium and the rare earth elements, where zirconium was used as a surrogate for the transuranic(TRU) elements. All the experiments were performed in a glove box filled with a argon gas. Li-Bi alloy was used as a reducing agent to reduce the high chemical activity of Li. The reductive extraction characteristics were examined using ICP, XRD and EPMA analysis. The reduction reaction was equilibrated within 3 hours after the Li addition. Three eutectic salt systems were compared and Zr was successfully separated from the rare earth elements in all the three salt systems.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.483-490
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• 2012

PWR spent fuels produced in the Republic of Korea are expected to be recycled by pyroprocess in the long term future. Even though pyroprocess waste amounts can be smaller than that of PWR spent fuel assembly in case of direct disposal, this process essentially will produce various and unique radioactive wastes. The goals of this article are to characterize these wastes, calculate the amount of wastes, design disposal systems for each waste and evaluate the radiation safety of each system by dose assessment. The absorbed dose results of the metal and ceramic waste for the engineering barrier system (EBS) showed $2.21{\times}10^{-2}$ Gy/h and $1.15{\times}10^{-2}$ Gy/h, which are lower than the recommended value of 1 Gy/h. These results confirmed that the newly proposed disposal systems have a safety margin for the radiation produced from each waste.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2889-2896
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• 2023

The waste acceptance criteria for heavy metals in mixed waste should be developed by reflecting the leaching behaviors that could highly depend on the repository design and environment surrounding the waste. The current standards widely used to evaluate the leaching characteristics of heavy metals would not be appropriate for the silo-type repository since they are developed for landfills, which are more common than a silo-type repository. This research aimed to explore the leaching behaviors of cementitious waste with Pb, Cd, and Sb metallic and oxide powders in an environment simulating a silo-type radioactive waste repository. The Toxicity Characteristic Leaching Procedure (TCLP) and the ANS 16.1 standard were employed with standard and two modified solutions: concrete-saturated deionized and underground water. The compositions and elemental distribution of leachates and specimens were analyzed using an inductively coupled plasma optical emission spectrometer (ICP-OES) and energy-dispersive X-ray spectroscopy combined with scanning electron microscopy (SEM-EDS). Lead and antimony demonstrated high leaching levels in the modified leaching solutions, while cadmium exhibited minimal leaching behavior and remained mainly within the cement matrix. The results emphasize the significance of understanding heavy metals' leaching behavior in the repository's geochemical environment, which could accelerate or mitigate the reaction.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.351-357
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• 2003

Transuranic elements such as Pu, Am and Cm in synthetic solution of spent nuclear fuel samples were determined by electrodeposition followed by alpha-spectrometry after separation using anion exchange and extraction chromatography in order to determine the transuranic elements in radwaste samples from nuclear power plants. Plutonium was separated by 12M HC1-0.1M HI as an eluent on anion exchange column. As a second step Am and Cm were separated in a group by DTPA-Lactic acid as the eluent on HDEHP coated column. The nuclides of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ separated were determined by alpha-spectrometry after electrodeposition in 0.1M $NaHSo_4$-0.53M $Na_2SO_4$buffer solution as an electrolyte. The recovery yields of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ were 83.8%, 85.2% and 86.3%, respectively, from the synthetic solution containing uranium and non-radioactive metal elements.