• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2329-2333
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• 2022

Base metal corrosion forms a significant issue during the chemical decontamination of the primary coolant loop in nuclear power plants as it is directly related to the economic and safety viability of decommissioning. In this technical note, potentiodynamic evaluations of several base metals (304 stainless steel, SA106 Grade B carbon steel, and alloy 600) were performed to determine their corrosion behavior during the hydrazine (N₂H₄)-based reductive ion decontamination (HyBRID) process. The results suggested that N₂H₄ protected the surface of the base metals in the HyBRID solution, which is primarily composed of H₂SO₄. The corrosion resistance of the carbon steel was further improved through the addition of CuSO₄ to the solution. The corrosion rate of carbon steel in the H₂SO₄-N₂H₄-CuSO₄ solution was lower than that exhibited in an oxalic acid solution, a commonly used reaction medium during commercial decontamination processes. These results indicate the superiority of the HyBRID process with respect to the base metal stability.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1140-1153
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• 2016

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Journal of the Korean Recycled Construction Resources Institute
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• v.6 no.2
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• pp.115-122
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• 2018

In this paper, the feasibility of recycling concrete waste as a method to reduce final disposal amount of wastes generated through decommissioning of nuclear power plant has been analyzed based on experimental results of existing literature. When recycled concrete waste was used as recycled aggregate, it was investigated through literature that the concrete strength decreased by 30~40% depending on the mixing ratio. It was also investigated that concrete with recycled aggregate can be used as a structural material when the quality of recycled aggregate is well managed since no significant problem was found. When recycled cement produced from concrete waste was used, the strength of concrete or mortar decreased considerably as the recycled cement content increased. Therefore, it can be concluded that concrete or mortar with recycled cement can be used as a filling material for final disposal of large radioactive waste rather than for structural use. This paper is expected to be useful for reduction on disposal volume and decommissioning cost for nuclear power plants such as Kori 1.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1289-1297
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• 2018

Preliminary surface soil Derived Concentration Guideline Levels (DCGLs) were derived conforming to the Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM) procedure for the site release and reuse of Kori Unit 1 in Korea. Based on the decommissioning experiences of the U.S. nuclear power plants, a suite of residual radionuclides was determined, and uncertainties contributed to the resultant dose by the input parameters were quantified via the sensitivity analysis of parameters. The peak of the mean dose was obtained via the probabilistic analysis of the RESRAD (RESidual RADioactivity)-ONSITE code. Consequently, $DCGL_w$ of Kori Unit 1 in accordance with two scenarios, industrial worker and residential farmer scenario, were derived and the results were compared respectively with other NPPs. It could be used as a basic guideline for establishing regulatory standards for reuse planning, designing the site characterization surveys and implementing final status survey (FSS).

• Journal of Environmental Impact Assessment
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• v.26 no.6
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• pp.479-494
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• 2017

After Fukushima incident, negative sentiment towards nuclear power has led to transition in policies that reduce the dependency on nuclear power in some countries. President Moon of Republic of Korea also announced a national plan of decommissioning retired nuclear power plants stage by stage. Therefore, nuclear power that once was considered the critical solution to energy security and climate change is now a limited option. This study aims to find an optimal energy mix in Korea's electricity system from 2016 through 2030 to combat climate change through energy transition with minimum cost. The study is divided into two different scenarios; energy transition and nuclear sustenance, to compare the total costs of the systems. Both scenarios show that electricity generated by wind technology increases from 2018 whereas that of photovoltaic(PV) increases from 2021. However, the total cost of the energy transition scenario was USD 4.7 billion more expensive than the nuclear sustenance scenario.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.517-536
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• 2020

The radionuclide inventory in radioactive waste from nuclear power plants should be determined to secure the safety of final repositories. As an alternative to time-consuming, labor-intensive, and destructive radiochemical analysis, the indirect scaling factor (SF) method has been used to determine the concentrations of difficult-to-measure radionuclides. Despite its long history, the original SF methodology remains almost unchanged and now needs to be improved for advanced SF implementation. Intense public attention and interest have been strongly directed to the reliability of the procedures and data regarding repository safety since the first operation of the low- and intermediate-level radioactive waste disposal facility in Gyeongju, Korea. In this review, statistical methodologies for SF implementation are described and evaluated to achieve reasonable and advanced decision-making. The first part of this review begins with an overview of the current status of the scaling factor method and global experiences, including some specific statistical issues associated with SF implementation. In addition, this review aims to extend the applicability of SF to the characterization of large quantities of waste from the decommissioning of nuclear facilities.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1483-1490
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• 2021

Many studies are undertaken into nuclear power plants (NPPs) in preparation for accidents exceeding design standards. In this paper, we analyze the applicability of various wireless communication technologies as accident countermeasures in different NPP environments. In particular, a commercial wireless communication module (WCM) is investigated by measuring leakage current and packet error rate (PER), which vary depending on the intensity of incident radiation on the module, by testing at a Co-60 gamma-ray irradiation facility. The experimental results show that the WCMs continued to operate after total doses of 940 and 1097 Gy, with PERs of 3.6% and 0.8%, when exposed to irradiation dose rates of 185 and 486 Gy/h, respectively. In short, the lower irradiation dose rate decreased the performance of WCMs more than the higher dose rate. In experiments comparing the two communication protocols of request/response and one-way, the WCMs survived up to 997 and 1177 Gy, with PERs of 2% and 0%, respectively. Since the request/response protocol uses both the transmitter and the receiver, while the one-way protocol uses only the transmitter, then the electronic system on the side of the receiver is more vulnerable to radiation effects. From our experiments, the tested module is expected to be used for design-based accidents (DBAs) of "Category A" type, and has confirmed the possibility of using wireless communication systems in NPPs.

• Fire Science and Engineering
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• v.20 no.3 s.63
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• pp.65-70
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• 2006

From the beginning of the construction stage, the fire protection regulation for the nuclear power plants conforms to the design requirements for the acquisition of the license permit. This regulation is based on the plant status of the normal operation, but it is not enough to be used as an application standard for fire protection at the transient mode of the plant and the outage time for refueling as well as for the plant decommissioning. While the advanced fire protection requirement that has been developed in America recently suggests the performance-based requirement and management rule applicable to the overall life time of the plant, it simply represents the conceptual application. It means that it can not be treated as appropriate standards because it does not deal with the qualitative and quantitative approach in specific ways. By the way, with the use of the performance-based fire risk analysis, the dynamic behavior of the heat and smoke at the fire compartment of the nuclear power plants can be analyzed and the thermal effect to the safety-related equipment and cables can be evaluated as well. At this paper, it suggests the ways to change the applicable fire protection regulations and the required evaluation items for the fire risk resulted from the plant configuration change with an intent to introduce the state-of-the-art quantitative fire risk analysis technology at the domestic nuclear power plants.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.44-51
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• 2023

Background: Aptamers are currently being used in various fields including medical treatments due to their characteristics of selectively binding to specific molecules. Due to their special characteristics, the aptamers are expected to be used to remove radionuclides from a large amount of liquid radioactive waste generated during the decommissioning of nuclear power plants. The radiological effects on the aptamers should be evaluated to ensure their integrity for the application of a radionuclide removal technique. Materials and Methods: In this study, Monte Carlo N-Particle transport code version 6 (MCNP6) and Monte Carlo damage simulation (MCDS) codes were employed to evaluate the radiological effects on the aptamers. MCNP6 was used to evaluate the secondary electron spectrum and the absorbed dose in a medium. MCDS was used to calculate the DNA damage by using the secondary electron spectrum and the absorbed dose. Binding experiments were conducted to indirectly verify the results derived by MCNP6 and MCDS calculations. Results and Discussion: Damage yields of about 5.00×10^-4 were calculated for 100 bp aptamer due to the radiation dose of 1 Gy. In experiments with radioactive materials, the results that the removal rate of the radioactive ⁶⁰Co by the aptamer is the same with the non-radioactive ⁵⁹Co prove the accuracy of the previous DNA damage calculation. Conclusion: The evaluation results suggest that only very small fraction of significant number of the aptamers will be damaged by the radioactive materials in the liquid radioactive waste.