• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2489-2497
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• 2023

Co and Eu impurities in the SSCs are nuclides that dominantly influence the neutron-induced radioactive inventory in metal and concrete radwastes (radioactive wastes) during NPP decommission. The impurity concentrations provided by NUREG/CR-3474 were used for the practical range of Co and Eu impurity concentrations to be applied to the code calculations. Metal structures near the core were evaluated to be ILW (intermediate-level waste) for the whole range of Co impurity concentration, so the boundary line between ILW and LLW (low-level waste) has no change for the whole concentration range provided by NUREG/CR-3474. Also, the boundary line between VLLW (very low-level waste) and CW (clearance waste) in the concrete shield could alter a little depending on the Eu impurity concentration within the range provided by NUREG/CR-3474. From this work, it is found that the concentration of material impurities of SSCs gives no critical impact on determining radwaste levels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.329-334
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• 2006

The depth distributions of total $^3H$ and $^{14}C$ activities were characterized for the activated shielding concrete from a decommissioning of KRR-2 using the commercially available tube furnace and a liquid scintillation counter. The correlation of measurement results between $^3H,\;^{14}C$ and gammer emitter was evaluated to apply for estimating radionuclide inventory of the concrete waste generated from decommissioning KRR-2. The detection limits for $^3H$ and $^{14}C$ are 0.048 and 0.028 Bq/g respectively. The specific activities of the $^3H$ and $^{14}C$ tend to decrease exponentially as the depth of the concrete becomes deeper from the surface. In addition, the $^3H$ and $^{14}C$ activities were in good correlation with the $^{60}CO$ activities analysed for the shielding concrete of KRR-2.

• Economic and Environmental Geology
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• v.54 no.2
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• pp.285-297
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• 2021

Globally, nuclear-decommissioning facilities have been increased in number, and thereby hundreds of thousands of wastes, such as concrete, soil, and metal, have been generated. For this reason, there have been numerous efforts and researches on the development of technology for volume reduction and recycling of solid radioactive wastes, and this study reviewed and examined thoroughly such previous studies. The waste concrete powder is rehydrated by other processes such as grinding and sintering, and the processes rendered aluminate (C3A), C4AF, C3S, and ��-C2S, which are the significant compounds controlling the hydration reaction of concrete and the compressive strength of the solidified matrix. The review of the previous studies confirmed that waste concretes could be used as recycling cement, but there remain problems with the decreasing strength of solidified matrix due to mingling with aggregates. There have been further efforts to improve the performance of recycling concrete via mixing with reactive agents using industrial by-products, such as blast furnace slag and fly ash. As a result, the compressive strength of the solidified matrix was proved to be enhanced. On the contrary, there have been few kinds of researches on manufacturing recycled concretes using soil wastes. Illite and zeolite in soil waste show the high adsorption capacity on radioactive nuclides, and they can be recycled as solidification agents. If the soil wastes are recycled as much as possible, the volume of wastes generated from the decommissioning of nuclear power plants (NPPs) is not only significantly reduced, but collateral benefits also are received because radioactive wastes are safely disposed of by solidification agents made from such soil wastes. Thus, it is required to study the production of non-sintered cement using clay minerals in soil wastes. This paper reviewed related domestic and foreign researches to consider the sustainable recycling of concrete waste from NPPs as recycling cement and utilizing clay minerals in soil waste to produce unsintered cement.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.202-207
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• 2012

The radioactivity inventory significantly affects all steps of decommissioning projects including planning, cost estimation, risk assessment, waste management and site remediation. The decommissioning project of the KRR-2 was completed in 2009 and a large amount of activated concrete waste was generated. The bioshield concrete, containing minute amount of impurity elements, was activated by neutron reaction during the operation of the reactor. A variety radionuclides was generated in the concrete, including $^3H$, $^{14}C$, $^{55}Fe$, $^{60}Co$ $^{63}Ni$, $^{134}Cs$, $^{152}Eu$ and $^{154}Eu$. In this paper, the comparison between the calculated results and previous measured results was carried out to estimate the inventory of the bioshield concrete of the KRR-2. The combined computer codes of MCNP5 and ORIGEN 2.1 for calculation of the distribution of neutron flux, cross-section and generation of radionuclides were used. The results were shown that 99.8% of the total radioactivity of $^3H$, $^{55}Fe$, $^{60}Co$ and $^{152}Eu$ in the bioshield concrete 12 years after shutdown. The effects on the variation of inventory were analysed depending on the operation periods and the cooling times in the bioshield concrete.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.157-167
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• 2020

Removing radioactive concrete is crucial in the decommissioning of nuclear power plants. However, this process generates radioactive aerosols, exposing workers to radiation. Although large amounts of radioactive concrete are generated during decommissioning, studies on the internal exposure of workers to radioactive aerosols generated from the cutting of radioactive concrete are very limited. In this study, therefore, we calculate the internal radiation doses of workers exposed to radioactive aerosols during activities such as drilling and cutting of radioactive concrete, using previous research data. The electrical-mobility-equivalent diameter measured in a previous study was converted to aerodynamic diameter using the Newton-Raphson method. Furthermore, the specific activity of each nuclide in radioactive concrete 10 years after nuclear power plants are shut down was calculated using the ORIGEN code. Eventually, we calculated the committed effective dose for each nuclide using the IMBA software. The maximum effective dose of ¹⁵²Eu constituted 83.09% of the total dose; moreover, the five highest-ranked elements (¹⁵²Eu, ¹⁵⁴Eu, ⁶⁰Co, ²³⁹Pu, ⁵⁵Fe) constituted 99.63%. Therefore, we postulate that these major elements could be measured first for rapid radiation exposure management of workers involved in decommissioning of nuclear power plants, even if all radioactive elements in concrete are not considered.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2019-2024
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• 2021

For evaluating the radiological safety of dismantled concrete, the process of disposal and recycling of the radioactive concrete generated during the dismantling of Kori Unit 1 is analyzed. Four scenarios are derived based on the analysis of the concrete recycling and disposal process, and the potential exposure to the workers and public during this process are calculated. VISIPLAN and RESRAD code are used for evaluating the dosages received by the workers and public in the following four scenarios: concrete inspection, transport of concrete by the truck driver, driving on a recycled concrete road, and public living near the landfilled concrete waste. Two worker exposure scenarios in the processing of concrete and two public exposure scenarios in recycling and disposal are considered; in all the scenarios, the exposure dose does not exceed the annual dose limit for each representative.

• 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 Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.203-212
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• 2019

Disposal nonconformity of radioactive wastes refers to radioactive wastes that need to be treated, solidified and packaged during operation or decommissioning of NPPs, and are typically exemplified by particulate radioactive wastes with dispersion characteristics. These wastes include the dried powders of concentrated wastes generated in the process of operating NPPs, slurry and sludge, various powdered wastes generated in the decommissioning process (crushed concrete, decontamination sludge, etc.), and fine radioactive soil, which is not easy to decontaminate. As these particulate wastes must be packaged so that they become non-dispersive, they are solidified with solidification agents such as cement and polymer. If they are treated using existing solidification methods, however, the volume of the final wastes will increase. This drawback may increase the disposal cost and reduce the acceptability of disposal sites. Accordingly, to solve these problems, this study investigates the pelletization of particulate radioactive wastes in order to reduce final waste volume.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.103-111
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• 2020

Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g^-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.

• The Korean Journal of Nuclear Medicine Technology
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• v.25 no.2
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• pp.41-47
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• 2021

Purpose There are many cyclotrons compared to the land area of the Republic of Korea. Because GMP certification is required and the nuclear medicine test does not apply for insurance, the number of examinations for nuclear medicine is decreasing. Therefore, there is a high probability of early decommissioning of the cyclotron. However, we do not unusually perform the radioactivation evaluation on concrete that can be classified as radioactive waste during the decommissioning of the cyclotron. In this study, we aim to confirm the radioactivation in the concrete surface using Handheld Radionuclide Identification Devices (RIDs). Materials and Methods Because there is no cyclotron being decommissioning in the Republic of Korea, it was impossible to perform the coring of concrete for radioactivation analysis. In this study, we used the KIRAMS-13 and analyzed the concrete surface in the target direction in the cyclotron room. After setting the target direction as the center, radionuclides were measured for about five months at thirty points with vertical and horizontal intervals of 30 cm. We used the RIIDEye(Detector: NaI(Tl) detector, manufacturer: Thermo) in this study and set the measurement time per point to one day (24 hours). Results Co-60 and Cs-137 were detected in some measurement points, and we confirmed the radioactivity of Co-60 detected at the most points. As a result, we found that the radioactivity of Co-60 was high in the diagonal direction (from the lower-left direction to the upper right direction) based on the center of the target. However, we think it is impossible to apply the corresponding results to all cyclotrons because we performed the study using only one cyclotron. Conclusion In thirty measurement points, we could confirm the radioactive nuclides and the relative radioactivity using the results of portable nuclides analyzer. Therefore, we expect that we can use the portable nuclides analyzer to select the coring position of concrete during the decommissioning of the cyclotron. Also, if we secure the radioactivation data for several years, we expect to make a more accurate estimate of radioactive waste during the preparation period of decommissioning of the cyclotron.