• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.267-277
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• 2016

For the decommissioning of a nuclear power plant, a site characterization and final status survey require a site-specific suite of radionuclides that could potentially still be present in the site during or after the decontamination processes. The United States Nuclear Regulatory Commission (U.S. NRC) requires a Decommissioning Technical Base Document (DTBD) along with a Site Characterization and Historical Site Assessment (HSA) from the utility for decommissioning to proceed. Both the DTBD and HSA are preliminary components of the Radiological Site Survey investigation process and should be included in the final License Termination Plan (LTP) for site release and reuse consideration from the U.S. NRC and the utility company. This study reviews the United States Nuclear Power Plants (U.S. NPPs) decommissioning cases and is especially focused on the methodologies used for determining a site-specific suite of radionuclides before and during the site characterization and final status surveys. In 2017, Kori-1 will be ready for decommissioning and related preparations are ongoing, this review will help Korea to prepare regulatory guidelines and give technical background for the safe and successful decommissioning of NPPs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.331-337
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• 2018

The decommissioning of nuclear power plants is generally executed in five steps, including preparation, decontamination, cutting/demolition, waste disposal and environmental restoration. So, for efficient decommissioning of nuclear power plants, worker safety, effects compared to cost, minimization of waste, possibility of reuse, etc., shall be considered. Worker safety and measurement technology shall be secured to exert optimal efficiency of nuclear power plant decommissioning work, for which accurate measurement technology for systems and devices is necessary. Typical In-Situ methods for decommissioning of nuclear plants are CZT, Gamma Camera and ISOCS. This study used ISOCS, which can be applied during the decommissioning of a nuclear power plant site without collecting representative samples, to take measurements of the S/G Water Chamber. To validate the measurement values, Microshield and the GEANT4 code was used as the actual method were used for modeling, respectively. The comparison showed a difference of $1.0{\times}10^1Bq$, which indicates that it will be possible to reduce errors due to the influence of radiation in the natural environment and the precision of modeling. Based on the research results of this paper, accuracy and reliability of measurement values will be analyzed and the applicability of the direct measurement method during the decommissioning of NPPs will be assessed.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.20-27
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• 2023

Background: As the number of nuclear facilities nearing their pre-determined design life increases, demand is increasing for technology and infrastructure related to the decommissioning and decontamination (D&D) process. It is necessary to consider the nature of the dismantling environment constantly changing and the worker doing new tasks. A method was studied that can calculate the effect of learning and the change in work time on the work process, according to the learning-forgetting curve model (LFCM). Materials and Methods: The LFCM was analyzed, and input values and scenarios were analyzed for substitution into the D&D process of a nuclear facility. Results and Discussion: The effectiveness and efficiency of the training were analyzed. It was calculated that skilled workers can receive a 16.9% less collective radiation dose than workers with only basic training. Conclusion: Using these research methods and models, it was possible to calculate the change in the efficiency of workers performing new tasks in the D&D process and the corresponding reduction in the work time and collective dose.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.84-85
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• 2009

New approaches for detecting, preventing and remedying environmental damage are important for protection of the environment. Procedures must be developed and implemented to reduce the amount of waste produced in chemical processes, to detect the presence and/or concentration of contaminants and decontaminate fouled environments. Contamination can be classified into three general types: airborne, surface and structural. The most dangerous type is airborne contamination, because of the opportunity for inhalation and ingestion. The second most dangerous type is surface contamination. Surface contamination can be transferred to workers by casual contact and if disturbed can easily be made airborne. The decontamination of the surface in the nuclear facilities has been widely studied with particular emphasis on small and large surfaces. The amount of wastes being produced during decommissioning of nuclear facilities is much higher than the total wastes cumulated during operation. And, the process of decommissioning has a strong possibility of personal's exposure and emission to environment of the radioactive contaminants, requiring through monitoring and estimation of radiation and radioactivity. So, it is important to monitor the radioactive contamination level of the nuclear facilities for the determination of the decontamination method, the establishment of the decommissioning planning, and the worker's safety. But it is very difficult to measure the surface contamination of the floor and wall in the highly contaminated facilities. In this study, the poly(styrene-ethyl acrylate) [poly(St-EA)] core-shell composite polymer for measurement of the radioactive contamination was synthesized by the method of emulsion polymerization. The morphology of the poly(St-EA) composite emulsion particle was core-shell structure, with polystyrene (PS)as the core and poly(ethyl acrylate) (PEA) as the shell. Core-shell polymers of styrene (St)/ethyl acrylate (EA) pair were prepared by sequential emulsion polymerization in the presence of sodium dodecyl sulfate (SOS) as an emulsifier using ammonium persulfate (APS) as an initiator. The polymer was made by impregnating organic scintillators, 2,5-diphenyloxazole (PPO) and 1,4-bis[5-phenyl-2-oxazol]benzene (POPOP). Related tests and analysis confirmed the success in synthesis of composite polymer. The products are characterized by IT-IR spectroscopy, TGA that were used, respectively, to show the structure, the thermal stability of the prepared polymer. Two-phase particles with a core-shell structure were obtained in experiments where the estimated glass transition temperature and the morphologies of emulsion particles. Radiation pollution level the detection about under using examined the beta rays. The morphology of the poly(St-EA) composite polymer synthesized by the method of emulsion polymerization was a core-shell structure, as shown in Fig. 1. Core-shell materials consist of a core structural domain covered by a shell domain. Clearly, the entire surface of PS core was covered by PEA. The inner region was a PS core and the outer region was a PEA shell. The particle size distribution showed similar in the range 350-360 nm.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.245-251
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• 2016

Soil can be contaminated by radioactive materials due to nuclide leakage following unexpected situations during the decommissioning of a nuclear power plant. Soil decontamination is necessary if contaminated land is to be reused for housing or industry. The present study classifies various soil remediation technologies into biological, physics/chemical and thermal treatment and analyzes their principles and treatment materials. Among these methods, this study selects technologies and categorizes the economics, applicability and technical characteristics of each technology into three levels of high, medium and low by weighting the various factors. Based on this analysis, the most applicable soil decontamination technology was identified.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.271-280
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• 2013

Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. And concrete waste was generated about 800 drums of 200 L. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. Thus, this research has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form to meet the requirements specified in disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10wt% as the optimized mixing ratio. Also, the compressive strength of cement waste form was satisfied that including fine powder up to maximum 40wt% in concrete debris wastes about 75%. As a result of scale-up test, the mixture of concrete waste, water, and cement is 75:10:15wt% meet the satisfied compressive strength because the free water increased with and increased in particle size.

• Journal of Radiation Protection and Research
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• v.24 no.2
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• pp.101-108
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• 1999

The influence of the relative humidity, the temperature and the velocity of supply air on evaporation rate has been studied with non-boiling forced evaporation system in order to treat very low level radioactive liquid wastes produced from the decontamination and decommissioning activities. Experimental data on the evaporation rate have been obtained with the divers variables and experimental equation of air velocity was also obtained by the correlation of those data. The decontamination factor of this system was also obtained by the experimental data from a simulated liquid waste containing Cs-137 radio isotope ; $DF=10^4$. Since the commercial system will be operated for the treatment of the very low level radioactive liquid waste produced from decontamination & decommissioning of TRIGA Mark-II&III research reactor, the environmental assessment has been conducted to improve the operational safety. Exposure dose rate for an individual member of general public was assessed, and it showed that it was very lower than individual dose limits. The release of radioactivity of radioisotope material (Cs-137) to the environment was assessed, and result showed that it was $4.637{\times}10^{-14}\;{\mu}Ci/cc$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.2
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• pp.227-235
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• 2024

After the permanent shut down of Kori Unit 1, various decommissioning activities will be implemented, including decontamination, segmentation, waste management, and site restoration. During the decommissioning period, waste management is among the most important activities to ensure that the process proceeds smoothly and within the expected timeframe. Furthermore, the radioactive waste generated during the operation should be sent to a disposal facility to complete the decommissioning project. Square and cylindrical concrete re-package drums were generated during the 1980s and 1990s. The square, containing boron concentrates, and cylindrical, containing spent resin, concrete re-package drums have been stored in a radioactive waste storage building. Homogeneous radioactive waste, including boron concentrates, spent resin, and sludge, should be solidified or packaged in high-integrity containers (HICs). This study investigates the sequential segmentation process for the separation of contaminated and non-contaminated regions, the re-packaging process of segmented or crushed cement-solidified boron concentrate, and re-packaging in HICs. The conceptual design evaluates the re-packaging plan for the segmented and crushed cement-solidified waste using HICs, which is acceptable in a disposal facility, and the quantity of generated HICs from the treatment process.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1210-1217
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• 2021

The analysis of ⁴¹Ca in concrete generated from the nuclear facilities decommissioning is critical for ensuring the safe management of radioactive waste. An analytical method for the determination of ⁴¹Ca in concrete is described. ⁴¹Ca is a neutron-activated long radionuclide, and hence, for accurate analysis, it is necessary to completely extract Ca from the concrete sample where it exists as the predominant element. The decomposition methods employed were the acid leaching, microwave digestion, and alkali fusion. A comparison of the results indicated that the alkali fusion is the most suitable way for the separation of Ca from the concrete sample. Several processes of hydroxide and carbonate precipitation were employed to separate ⁴¹Ca from interferences. The method relies on the differences in the solubility of the generated products. The behavior of Ca and the interfering elements such as Fe, Ni, Co, Eu, Ba, and Sr is examined at each separation step. The purified ⁴¹Ca was measured by a liquid scintillation counter, and the quench curve and counting efficiency were determined by using a certified reference material of known ⁴¹Ca activity. The recoveries in this study ranged from 56 to 68%, and the minimum detectable activity was 50 mBq g^-1 with 0.5 g of concrete sample.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.45-63
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• 2017

The oldest commercial reactor in South Korea, Kori-1 Nuclear Power Plant (NPP), will be shut down in 2017. Proper treatment for decommissioning wastes is one of the key factors to decommission a plant successfully. Particularly important is the recycling of clearance level or very low level radioactively contaminated metallic wastes, which contributes to waste minimization and the reduction of disposal volume. The aim of this study is to introduce a conceptual design of a recycle system and to evaluate the doses incurred through defined work flows. The various architecture diagrams were organized to define operational procedures and tasks. Potential exposure scenarios were selected in accordance with the recycle system, and the doses were evaluated with the RESRAD-RECYCLE computer code. By using this tool, the important scenarios and radionuclides as well as impacts of radionuclide characteristics and partitioning factors are analyzed. Moreover, dose analysis can be used to provide information on the necessary decontamination, radiation protection process, and allowable concentration limits for exposure scenarios.