• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4181-4190
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• 2023

Nuclear power plants are subjected to various processes during decommissioning, including cutting, decontamination, disposal, and treatment. The cutting of massive bio-shields is a significant step in the decommissioning process. Cutting is performed near the target structure, and during this process, workers are exposed to potential radioactive elements. However, studies considering worker exposure management during such cutting operations are limited. Furthermore, dismantling a nuclear power plant under certain circumstances may result in the unnecessary radiation exposure of workers and an increase in secondary waste generation. In this study, a cutting scenario was formulated considering the bio-shield as a representative structure. The specifications of a standard South Korean radioactive waste disposal drum were used as the basic conditions. Additionally, we explored the hot-to-cold and cold-to-hot methods, with and without the application of polishing during decontamination. For evaluating various scenarios, different cutting time points up to 30 years after permanent shutdown were considered, and cutting speeds of 1-10nullm²/h were applied to account for the variability and uncertainty attributable to the design output and specifications. The obtained results provide fundamental guidelines for establishing cutting methods suitable for large structures.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.4
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• pp.469-477
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• 2021

Safe storage of spent nuclear fuel in deep underground repositories necessitates an understanding of the long-term alteration of metal canisters and buffer materials. A small-scale laboratory alteration test was performed on metal (Cu or Fe) chips embedded in compacted bentonite blocks placed in anaerobic water for 1 year. Lactate, sulfate, and bacteria were separately added to the water to promote biochemical reactions in the system. The bentonite blocks immersed in the water were dismantled after 1 year, showing that their alteration was insignificant. However, the Cu chip exhibited some microscopic etch pits on its surface, wherein a slight sulfur component was detected. Overall, the Fe chip was more corroded than the Cu chip under the same conditions. The secondary phase of the Fe chip was locally found as carbonate materials, such as siderite (FeCO₃) and calcite ((Ca, Fe)CO₃). These secondary products can imply that the local carbonate occurrence on the Fe chip may be initiated and developed by an evolution (alteration) of bentonite and a diffusive provision of biogenic CO₂ gas. These laboratory scale results suggest that the actual long-term alteration of metal canisters/bentonite blocks in the engineered barrier could be possible by microbial activities.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.491-496
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• 2003

In Korea, the dose limits to the public were reduced according to ICRP-60 recommendations. The secondary quantities, Effluent Concentration Limits (ECLs) were derived and enacted to Korean Atomic Laws based on ICRP-60 recommendations. The Korea atomic laws require assurance that radioactive materials within gaseous effluents do not exceed dose limits and ECLs. This simply means that any effluent that would possibly contain radioactivity must be monitored. There are various methods to monitor the radioactivity of effluent monitor to satisfy the dose limits and the ECLs for gaseous effluents. The many factors (safety margin) should be considered in determining of the setpoint of effluent monitor, following these limits. In this study, we studied the determination method of alarm setpoint for gaseous effluent Radiation Monitoring Systems using dose factors considered the main pathway of radionuclides to compare the preceding determination method of alarm setpoint for gaseous effluent RMSs using dose assessment program considered all the practicable pathways of radionuclides.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.101-112
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• 2016

This study investigated the removal of Sr, which was one of the high radioactive nuclides, by adsorption with Barium (Ba) impregnated 4A zeolite (BaA) from high-radioactive seawater waste (HSW). Adsorption of Sr by BaA (BaA-Sr), in the impregnated Ba concentration of above 20.2wt%, was decreased by increasing the impregnated Ba concentration, and the impregnated Ba concentration was suitable at 20.2wt%. The BaA-Sr adsorption was added to the co-precipitation of Sr with $BaSO_4$ precipitation in the adsorption of Sr by 4A (4A-Sr) within BaA. Thus, it was possible to remove Sr more than 99% at m/V (adsorbent weight/solution volume)=5 g/L for BaA and m/V >20 g/L for 4A, respectively, in the Sr concentration of less than 0.2 mg/L (actual concentration level of Sr in HSW). It shows that BaA-Sr adsorption is better than 4A-Sr adsorption in for the removal capacity of Sr per unit gram of adsorbent, and the reduction of the secondary solid waste generation (spent adsorbent etc.). Also, BaA-Sr adsorption was more excellent removal capacity of Sr in the seawater waste than distilled water. Therefore, it seems to be effective for the direct removal of Sr from HSW. On the other hand, the adsorption of Cs by BaA (BaA-Cs) was mainly performed by 4A within BaA. Accordingly, it seems to be little effect of impregnated Ba into BaA. Meanwhile, BaA-Sr adsorption kinetics could be expressed the pseudo-second order rate equation. By increasing the initial Sr concentrations and the ratios of V/m, the adsorption rate constants ($k_2$) were decreased, but the equilibrium adsorption capacities ($q_e$) were increasing. However, with increasing the temperature of solution, $k_2$ was conversely increased, and $q_e$ was decreased. The activation energy of BaA-Sr adsorption was 38 kJ/mol. Thus, the chemical adsorption seems to be dominant rather than physical adsorption, although it is not a chemisorption with strong bonding form.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3293-3298
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• 2022

Removing radioactive contaminated metal materials is a vital task during the decommissioning of nuclear power plants to reduce the cost of the post-dismantling process. The laser decontamination technique has been recognized as a key tool for a successful dismantling process as it enables a remote operation in radioactive facilities. It also minimizes exposure of workers to hazardous materials and reduces secondary waste, increasing the environmental friendless of the post-dismantling processing. In this work, we present a thorough and efficient laser decontamination approach using a single-mode continuous-wave (CW) laser. We subjected stainless steels to a surface-removal process that repetitively exposes the laser to a confined region of ~75 ㎛ at a high scanning rate of 10 m/s. We evaluate the decontamination performance by measuring the removal depth with a 3D scanning microscope and further investigate optimal removal conditions given practical parameters such as the laser power and scan properties. We successfully removed the metal surface to a depth of more than 40 ㎛ with laser power of 300 W and ten scans, showing the potential to achieve an extremely high DF more than 1000 by simply increasing the number of scans and the laser power for the decontamination of primary circuits.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.220-237
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• 2004

China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basic strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan(2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before 2010 respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor(CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached the top namely 57m above the ground. More than one hundred components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000$m^2$ floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which is only under consideration up to now. Some important technical selections have been settled, but its design has not yet started.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.337-346
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• 2020

Two commonly used ceramics in molten salt research are alumina and mullite. The two ceramics were exposed to a combination of rare earth chlorides (YCl₃, SmCl₃, NdCl₃, PrCl₃, and CeCl₃; each rare earth chloride of 1.8 weight percent) in LiCl-KCl at 773 K for approximately 13 days. Scanning electron microscopy with wave dispersion spectra was utilized to investigate a formation layer or deposition of rare earths onto the ceramic. Only the major constituents of the ceramics (Al, Si, and O₂) were observed during the wave dispersion spectra. X-ray fluorescence was used as well to determine concentration changes in the molten salt as a function of ceramic exposure time. This study shows no evidence of ionic exchange or layer formation between the ceramics and molten chloride salt mixture. There are signs of surface tension effects of molten salt moving out of the tantalum crucible into secondary containment.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2736-2747
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• 2022

The operations in the design lifecycle of a nuclear power plant targeted to be decommissioned lead to neutron activation. Operations in the decommissioning process include cutting, decontamination, disposal, and processing. Among these, cutting is done close to the target material, and thus workers are exposed to radiation. As there are only a few studies on pressurizers, there arises the need for further research to assess the radiation exposure dose. This study obtained the specifications of the AP1000 pressurizer of Westinghouse and the distribution of radionuclide inventory of a pressurizer in a pressurised water reactor for evaluation based on literature studies. A cutting scenario was created to develop an optimal method so that the cut pieces fill a radioactive solid waste drum with dimensions 0.571 m × 0.834 m. The estimated exposure dose, estimated using the tool VISIPLAN SW, in terms of the decontamination factor (DF) ranged from DF-0 to DF-100, indicating that DF-90 and DF-100 meet the ICRP recommendation on exposure dose 0.0057 mSv/h. At the end of the study, although flame cutting was considered the most efficient method in terms of cutting speed, laser cutting was the most reasonable one in terms of the financial aspects and secondary waste.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2918-2925
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• 2021

The efficiency of decontamination of model spent ion exchange resins, contaminated with magnetite and hematite, with mineral acid solutions, and using electro-decontamination, was evaluated. It has been shown that effective hematite dissolution occurs in concentrated mineral acid solutions. However, the use of direct current increases the decontamination efficiency of spent ion exchange resins contaminated with hematite. It is determined that with increasing voltage and acid concentration, the dissolution efficiency of hematite deposits increases and can exceed 99%. It has been shown that hematite dissolution is accompanied by secondary adsorption of radionuclides due to ion exchange, which can be removed with sodium nitrate solutions.

• Journal of Electrochemical Science and Technology
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• v.10 no.2
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• pp.123-130
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• 2019

We investigated the electrochemical behavior of an electrode coated with a nickel hexacyanoferrate/graphene oxide (NiPB/GO) composite to evaluate its potential use for the electrochemical separation of radioactive Cs as a promising approach for reducing secondary Cs waste after decontamination. The NiPB/GO-modified electrode showed electrochemically switched ion exchange capability with excellent selectivity for Cs over other alkali metals. Furthermore, the repetitive ion insertion and desertion test for assessing the electrode stability showed that the electrochemical ion exchange capacity of the NiPB/GO-modified electrode increased further with potential cycling in 1 M of $NaNO_3$. In particular, this electrochemical treatment enhanced Cs uptake by nearly two times compared to that of NiPB/GO and still retained the ion selectivity of NiPB, suggesting that the electrochemically treated NiPB/GO composite shows promise for nuclear wastewater treatment.