• Journal of Radiation Protection and Research
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• v.39 no.2
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• pp.89-95
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• 2014

In potential accident consequence assessments for the licensing approval of LWRs, the ground deposition of radionuclides released into the environment is not allowed into the models, as recommended in the U. S. Nuclear Regulatory Commission's regulatory guide. Meanwhile, it is allowed into the assessment models for the licensing approval of PHWRs with consideration of more detailed physical processes of radionuclides in the atmosphere. Under these backgrounds, importance of exposure dose by ground deposition was quantitatively evaluated and comprehensively discussed. For potential accidental releases of $^{137}Cs$ and $^{131}I$, total exposure doses were more conservative in case of without consideration of ground deposition than in case of with its consideration. It was because of that the depletion of air concentration resulting from ground deposition is more influential in the contribution to total exposure doses than additional doses from contaminated ground. The exposure doses by the inhalation of contaminated air showed the contribution of more than 90% in total exposure doses, depending on atmospheric stability, release period of radionuclides and distance from a release point. The exposure doses from contaminated ground showed less than 10% at most in contribution of total exposure doses. The ratios of total exposure doses in case of with consideration of deposition to without its consideration for $^{131}I$ were distinct than those for $^{137}Cs$. As the atmosphere is more stable, release duration of radionuclides is longer, distance from a release point is longer, it was more distinct.

• Nuclear Medicine and Molecular Imaging
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• v.40 no.2
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• pp.90-95
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• 2006

Neuroendocrine tumors (NETs) consist of a heterogeneous group of tumors that are able to uptake neuroamine and/or specific receptors, such as somatostatin receptors, which can play important roles of the localization and treatment of these tumors. When considering therapy with radionuclides, the best radioligand should be carefully investigated. $^{131}I$-MIBG and beta-particle emitter labeled somatostatin analogs are well established radionuclide therapy modalities for NETs. $^{111}In,\;^{90}Y\;and\;^{177}Lu$ radiolabeled somatostatin analogues have been used for treatment of NETs. Further, radionuclide therapy modalities, for example, radioimmunotherapy, radiolabeled peptides such as minigastrin are currently under development and in different phases of clinical investigation. for all radionuclides used for therapy, long-term and survival statistics are not yet available and only partial tumour responses have been obtained using $^{131}I$-MIBG and $^{111}In$-octreotide. Experimental results using $^{90}Y$-DOTA-lanreotide as well as $^{90}Y-DOTA-D-Phe1-Tyr^3-octreotide$ and/or $^{177}Lu-DOTA-Tyr^3-octreotate$ have indicated the possible clinical potential of radionuclides receptor-targeted radiotherapy it may be hoped that the efficacy of radionuclide therapy will be improved by co-administration of chemotherapeutic drugs whose antitumoral properties may be synergistic with that of irradiation.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.556-561
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• 2017

Serving as an excellent adsorbent and inorganic ion exchanger in the water purification field, zeolite 4A has in this work presented a strong capability for purifying radioactive waste, such as $Sr^{2+}$, $Cs^+$, and $Co^{2+}$ in water. During the processes of decontamination and decommissioning of suspended solids and organics in low-level radioactive wastewater, the purification performance of zeolite 4A has been studied. Under ambient temperature and neutral condition, zeolite 4A absorbed simulated radionuclides such as $Sr^{2+}$, $Cs^+$, and $Co^{2+}$ with an absorption rate of almost 90%. Additionally, in alkaline condition, the adsorption percentage even approached 98.7%. After conducting research on suspended solids and organics of zeolite 4A for the treatment of radionuclides, it was found that the suspended clay was conducive to absorption, whereas the absorption of organics in solution was determined by the species of radionuclides and organics. Therefore, zeolite 4A has considerable potential in the treatment of radioactive wastewater.

• Journal of Radiation Protection and Research
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• v.49 no.1
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• pp.1-18
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• 2024

Exponential growth has been observed in nuclear medicine procedures worldwide in the past decades. The considerable increase is attributed to the advance of positron emission tomography and single photon emission computed tomography, as well as the introduction of new radiopharmaceuticals. Although nuclear medicine procedures provide undisputable diagnostic and therapeutic benefits to patients, the substantial increase in radiation exposure to nuclear medicine patients raises concerns about potential adverse health effects and calls for the urgent need to monitor exposure levels. In the current article, model-based internal dosimetry methods were reviewed, focusing on Medical Internal Radiation Dose (MIRD) formalism, biokinetic data, human anatomy models (stylized, voxel, and hybrid computational human phantoms), and energy spectrum data of radionuclides. Key results from many articles on nuclear medicine dosimetry and comparisons of dosimetry quantities based on different types of human anatomy models were summarized. Key characteristics of seven model-based dose calculation tools were tabulated and discussed, including dose quantities, computational human phantoms used for dose calculations, decay data for radionuclides, biokinetic data, and user interface. Lastly, future research needs in nuclear medicine dosimetry were discussed. Model-based internal dosimetry methods were reviewed focusing on MIRD formalism, biokinetic data, human anatomy models, and energy spectrum data of radionuclides. Future research should focus on updating biokinetic data, revising energy transfer quantities for alimentary and gastrointestinal tracts, accounting for body size in nuclear medicine dosimetry, and recalculating dose coefficients based on the latest biokinetic and energy transfer data.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.88-94
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• 2020

Background: The International Commission on Radiological Protection (ICRP) has recently published report series on the occupational intakes of radionuclides (OIR) for internal dosimetry of radiation workers. In this study, the optimized monitoring program including the monitoring interval and the minimum detectable activity (MDA) of major radionuclides was suggested to perform the routine individual monitoring of internal exposure based on the ICRP OIR. Materials and Methods: The derived recording levels and the critical monitoring quantities were reviewed from international standards or guidelines by the International Atomic Energy Agency (IAEA), the International Organization for Standardization (ISO), and the European Radiation Dosimetry Group (EURADOS). The OIR data viewer provided by ICRP was used to evaluate the monitoring intervals and the MDA, which are derived from the reference bioassay functions and the dose coefficients. Results and Discussion: The optimal monitoring intervals were determined taking account of two requirement conditions on the potential intake underestimation and the MDA values. The MDA requirement values of the selected radionuclides were calculated based on the committed effective dose from 0.1 mSv to 5 mSv. The optimized routine individual monitoring program was suggested including the optimal monitoring intervals and the MDA requirements. The optimal MDA values were evaluated based on the committed effective dose of 0.1 mSv. However, the MDA can be adjusted considering the practical operation of the routine individual monitoring program in the nuclear facilities. Conclusion: The monitoring intervals and the MDA as crucial factors for the routine monitoring were described to suggest the optimized routine individual monitoring program of the occupational intakes. Further study on the alpha/beta-emitting radionuclides as well as short lived gamma-emitting nuclides will be necessary in the future.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.879-884
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• 1995

The results of performance assessment analyses have shown that the long-lived radionuclides such as I-129 control the potential individual dose impact to the public. I-129 is difficult-to-measure(DTM) in low-level waste because it is non-gamma emitting radionuclides and exists at extremely low concentrations in radioactive waste generated by nuclear reactors. In this study, computer modeling technique to predict release rate of I-129 is developed to provide another tools far performance assessment of land disposal facilities and characteristics of radwaste. Model suggested in this study will give conservative values of I-129 release rate far determination of radwaste characteristics. More detailed approach is implemented to account for release conditions of fuel source-nuclides. 1-131 concentration measured from reactor coolant and released fraction from tramp fuel have dominant roles in calculating release rate of I-129 with fuel defect conditions.

• Journal of Radiation Protection and Research
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• v.31 no.3
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• pp.135-140
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• 2006

Intakes of radionuclides through both inhalation and ingestion pathways may occur particularly in an incident involving unsealed radionuclides. If one assume only one intake path in this case, which is usual in routine monitoring, a significant error in the evaluated committed effective dose($E_{50}$) may result. In order to demonstrate the potential errors, variations of the resulting committed effective doses were analyzed for different fractions of the inhaled activities to the total intake of $^{241}Am$. Simulated bioassav measurements for the lungs, urine and feces were generated based on the biokinetic model and data of the radionuclide, 5 ${\mu}m$ AMAD and absorption type M for inhalation, for various inhalation fractions. The potential errors in $E_{50}$ due to the assumption of one intake path were in the range from -100% to as large as +34,000% when the bioassays were made 3 days after the intakes. Larger errors are expected when only the feces assay is applied while inhalation intake exists. A strategy which employs two types of bioassay was proposed to reduce the error caused by a misjudgement of the intake path.

• Proceedings of the KSEEG Conference
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• 2002.06a
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• pp.81-100
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• 2002

While most of regulatory communities in abroad recognize ' 'natural attenuation " to include degradation, dispersion, dilution, sorption (including precipitation and transformation), and volatilization as governing Processes, regulators prefer "degradation" because this mechanism destroys the contaminant of concern. Unfortunately, true degradation only applies to organic contaminants and short- lived radionuclides, and leaves most metals and long-lived radionuclides. The natural attenuation Processes may reduce the potential risk Posed by site contaminants in three ways: (i)contaminants could be converted to a less toxic form througy destructive processes such as biodegradation or abiotic transformations; (ii) potential exposure levels may be reduced by lowering concentrations (dilution and dispersion); and (iii) contaminant mobility and bioavailability may be reduced by sorption to geomedia. In this review, authors will focus will focul on "sorption" among the natural attenuation processes of hazardous inorganic contaminants including radionuclides. Note though that sorption and transformation processes of inorganic contaminants in the natural setting could be influenced by biotic activities but our discussion would limit only to geochemical reactions involved in the natural attenuation. All of the geochemical reactions have been studied in-depth by numerous researchers for many years to understand "retardation" process of contaminants in the geomedia. The most common approach for estimating retardation is the determination of distrubution coefficiendts ($K_{d}$) of contaminants using parametric or mechanistic models. As typocally used in fate and contaminant transport calculations such as predictive models of the natural attenuation, the $K_{d}$ is defined as the ratio of the contaminant concentration in the surrounding aqueous solution when the system is at equilibrium. Unfortunately, generic or default $K_{d}$ values can result in significant error when used to predict contaminant migration rate and to select a site remediation alternative. Thus, to input the best $K_{d}$ value in the contaminant transport model, it is essential that important geochemical processes affecting the transport should be identified and understood. Precipitation/dissolution and adsorption/desorption are considered the most important geochemical processes affecting the interaction of inorganic and radionuclide contaminants with geomedia at the near and far field, respectively. Most of contaminants to be discussed in this presentation are relatively immobile, i.e., have very high $K_{d}$ values under natural geochemical environments. Unfortunately, the obvious containment in a source area may not be good enough to qualify as monitored natural attenuation site unless owner demonstrate the efficacy if institutional controls that were put in place to protect potential receptors. In this view, natural attenuation as a remedial alternative for some of sites contaminated by hazardous-inorganic components is regulatory and public acceptance issues rather than scientific issue.

• Journal of Radiation Protection and Research
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• v.35 no.1
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• pp.12-20
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• 2010

In a primary system at nuclear power plants (NPPs), various radionuclides including fission products and corrosion products are generated due to the complex water conditions. Particularly, $^3H,\;^{14}C,\;^{58}Co,\;^{60}Co,\;^{137}Cs,\;and^{131}I$ are important radionuclides in respect of dose assessment for radiation workers and management of radioactive effluents. In this paper, the dominant contributors of radiation exposure for radiation workers and the member of public adjacent to NPPs were reviewed and the process of dose assessment attributable to those contributors were introduced. Furthermore, the analysis for some examples of radiation exposure to radiation workers and the public during the NPP operation was carried out. This analysis included the notable precedents of internal radiation exposure and contamination of demineralized water occurred in Korean NPPs. Particularly, the potential issue about the dose assessment of tritium and carbon-14 was also reviewed in this paper.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1098-1108
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• 2019

Korea's first commercial nuclear power plant at Kori site was permanently shut down in 2017 and is currently in transition stage. Preparatory activities for decommissioning such as historical site assessment, characterization, and dismantling design are being actively carried out for successful D&D (Dismantling and Decontamination) at Kori site. The ultimate goal of decommissioning will be to ensure the safety of workers and residents that may arise during the decommissioning of nuclear facilities and, thereby finally returning the site to its original status in accordance with the release criteria. Upon completion of decommissioning, the resident's safety at a site released will be assessed from the evaluation of dose caused by radionuclides expected to be present or detected at the site. Although the U.S. commercial nuclear power plants with decommissioning experience use different site release criteria, most of them are 0.25 mSv/y. In Korea, both the unrestricted and restricted release criteria have been set to 0.1 mSv/y by the Nuclear Safety and Security Commission. However, since the dose is difficult to measure, measurable concentration guideline levels for residual radionuclides that result in dose equivalent to the site release criteria should be derived. For this derivation, site reuse scenario, selection of potential radionuclides, and systematic methodology should be developed in planning stage of Kori site decommissioning. In this paper, for calculation of a preliminary site-specific Derived Concentration Guideline Levels (DCGLs) for the Nuclear Power Plant site, a novel approach has been developed which can fully reflect practical reuse plans of the Kori site by taking into account multiple site reuse scenarios sequentially, thereby striking a remarkable distinction with conventional approaches which considers only a single site scenario.