• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.171-178
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• 2014

After the expiration of operating license of a plant, all infrastructures within the plant must be safely dismantled to the point that it no longer requires measures for radiation protection. Despite the fact that Kori 1 and Wolsong 1 are close to the expiration of their operating license, sufficient technologies for radiological characterization, decontamination and dismantling is still under development. The purpose of this study is to develop one of methods for radiological inventory assessment on measuring object by using direct measure of large component with In-Situ measurement technique. Radiological inventory was assessed by analyzing nuclide using portable gamma spectroscopy without dismantling reactor head, and the result of direct measurement was supplemented by performing indirect measurement. Radiochemical analysis were performed on surface contamination samples as well. During the study, radiological inventory of reactor vessel calculated expanding the result. Based on the result and the radioactivity variation of each radionuclides time frame for decommissioning can be decided. Thus, it is expected that during the decommissioning of plants, the result of this study will contribute to the reduction of radiation exposure to workers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.43-49
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• 2020

An engineered barrier system (EBS) for the geological disposal of high-level radioactive waste (HLW) consists of a disposal canister packed with spent fuel, buffer material, backfill material, and gap-filling material. The buffer material fills the space between the canister and the near-field rock, thus serving to restrain the release of radionuclides and protect the canister from groundwater penetration. Furthermore, as significant amounts of heat energy are released from the canister to the surrounding rock, the thermal conductivity of the buffer plays an important role in maintaining the safety of the entire disposal system. Therefore, given the high levels of heat released from disposal canisters, this study measured the thermal conductivities of compacted bentonite buffers from Gyeongju under temperature variations ranging 25 to 80~90℃. There was a 5~20% increase in thermal conductivity as the temperature increased, and the temperature effect increased as the degree of saturation increased.

• Journal of Sensor Science and Technology
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• v.21 no.5
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• pp.367-373
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• 2012

We fabricated a fiber-optic alpha/beta detector, which is composed of a sensing probe, a plastic optical fiber, a photomultiplier tube, and a multichannel analyzer, to obtain the energy spectra of radioactive isotopes. As inorganic scintillators of a sensing probe, a ZnS(Ag) film was coupled with a $CaF_2$(Eu) crystal for alpha and beta spectroscopy. In this study, $^{210}Po$ and $^{90}Sr$ were used as alpha and beta sources, respectively, and we measured the radiation energy spectra using a fiber-optic alpha/beta detector to identify alpha and beta emitting radionuclides for nuclear medicine application. Also, the variations of energy spectrum were obtained according to the length of plastic optical fiber.

• Journal of Environmental Impact Assessment
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• v.22 no.1
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• pp.89-97
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• 2013

RADCONS Ver. 1.0 (RADiological CONSequence Assessment Program) was developed for radiological risk assessment in this study. A Gaussian plume model was used to analyze the fate and transport of radionuclides released into the air in case of accidents. Both single meterological data and time series meterological data can be used in RADCONS. To assess the radiological risk of the early phase after an accident, ED (Effective Dose) estimated by both deterministic and probabilistic approaches are presented. These EDs by deterministic and probabilistic will be helpful to efficient decision making for decision makers. External doses from deposited materials by time are presented for quantifying the effects of mid and late phases of an accident. A radiological risk assessment was conducted using RADCONS for an accident scenario of 1 Ci of Cs-137. The maximum of ED for radii of 1,000 meters from the accident point was 8.51E-4 mSv. After Monte-Carlo simulation, considering the uncertainty of the breathing rate and dispersion parameters, the average ED was 8.49E-4, and the 95 percentile was 1.10E-3. A data base of the dose coefficients and a sampling module of the meteorological data will be modified to improve the user's convenience in the next version.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.8-15
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• 1989

After administration of $^{99m}TC,\;^{67}Ga,\;^{131}I,\;^{32}P\;and\;^{238}U(UO_2(NO_2)_3{\cdot}6H_2O)$, in male rats, toxic effects were examined by determining the biological materials in blood. Dosages of radio-nuclides injected are based on the amounts routinely administered to patients, and concentrations of BUN, creatinine, SGOT, SGPT and $PGE_2$ in plasma are determinated as indices to the biochemical response. No increase of creatinine was observed after injection of $^{99m}TC$. Concentrations of BUN, SGOT and $PGE_2$ were not significantly increased in comparison with before-administration. Administration of $^{67}Ga,\;^{131}I\;and\;^{32}P$, did not significantly change BUN, SGPT and SGOT, but largely increased $PGE_2$ than control levels. Besides, $^{238}U$ Showed the most severe toxicity. From the above results, we suggest that the determination of $PGE_2$ in plasma can be used as an index in case of elvaluating the effects of radiation toxicity by nuclides used in nuclear medicine.

• Journal of Radiation Protection and Research
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• v.30 no.3
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• pp.149-159
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• 2005

The purpose of this paper is to present the general attributes of the current IAEA programs and their prospects in the arena of protection of the marine and terrestrial environment, thus responding to a need to attain a consolidated understanding of the Agency's programs for an effective and efficient deployment of the respective national R&D projects in Korea. In addition, the considerable and beneficial benefits from a participation in the IAEA programs have been analyzed and their immediate relevance has been emphasized. A strategy for the enhancement of an active participation in the program and its efficient implementation has also been established. It is expected that the suggested recommendations such as the long term strategy and the relevant guidelines will be helpful in establishing a nuclear policy for the further development of the international cooperative projects in the future.

• Journal of Radiation Protection and Research
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• v.25 no.1
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• pp.3-9
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• 2000

The Korea Atomic Energy Research Institute acquired the Lawrence Livermore National Laboratory phantom for calibration of germanium detectors used for in vivo measurement of radionuclides deposited in human lungs. The manufacturer inscribed concentric circles as a phoswich detector positioning guides on the phantom's torso plate and its overlay plates, and provided the effective thickness of the chest wall for each plate measured at locations over the circles. However, since the germanium detectors are of different sizes, the areas considered for phoswich detectors were no. longer applicable for the locations of the germanium detectors on the phantom. Therefore, we re-evaluated the effective thickness of the phantom to determine if the manufacturer' s data are valid for germanium detectors in use for in vivo lung counting or if new values must be implemented. Differences no more than 3% in effective thickness were found between the germanium detector regions to be used at the Korea Atomic Energy Research Institute and the phoswich detector regions prescribed by the manufacturer.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.545-552
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• 2002

We developed an ultra low background gamma ray spectrometer particularly suitable for experiment which require lower detection limit. The background of a germanium spectrometer is suppressed by applying active and passive shielding technique at the same time. The active shielding devices consist of plastic scintillating plates of 50 mm thick and anti-coincidence electronic system. The shielding is made of 150 mm thick walls of very low activity lead,20 mm with activity of <10 Bq/kg and 130 mm with activity of <50 Bq/kg. The observed background count rates are 1.2 $s^{-1}$ and 0.36 $s^{-1}$ without and with the active shielding, respectively, overall the energy regions from 30 keV to 3 MeV The cosmic ray induced background is suppressed by a rate of 0.8 $s^{-1}$ at the present work. The detection efficiency curve necessary to obtain the radioactivity of environmental samples has been precisely determined on the energy regions from 80 to 2000 keV with a 10$^3$ ml marinelli beaker sample, consisting of the calibrated radionuclides $^{109}$ Cd, $^{57}$ CO, $^{139}$ Ce, $^{203}$ Hg, $^{113}$ Sn, $^{85}$Sr, $^{137}$ Cs, $^{60}$ Co and $^{88}$ Y. Virtues Of the method are demonstrated by measuring the activity of $^{137}$ Cs contained in the powdered milk.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.253-262
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• 2004

Construction of tunnels in a deep crystalline host rock for a potential High-Level Radioactive Waste(HLW) repository inevitably generates an excavation disturbed zone (EDZ). There have been a series of debates on whether a permeability in an EDZ increases or not and what would be the maximum depth of an EDZ. Recent studies show mixed opinions on permeability. However, there has been an international consensus on the thickness of an EDZ; 30 cm for TBM and 1 meter for controlled blast. One of the impacts of an EDZ is on determining the distance between adjacent deposition holes. The void gap by the excavation hinders relaxation of temperature profiles so that the current Korean reference designing distance between holes should be stretched out more to keep the maximum temperature in a buffer region below 100 degrees Celsius. The other impact of an EDZ is on the long-term post closure radiological safety. To estimate the impact, the reference scenario, the well scenario, is chosen. Released nuclides diffuse through a bentonite buffer region experiencing strong sorption and reach a fracture surrounded by a porous medium. Inside a fractured porous region, radionuclides migrate by advection and dispersion with matrix diffusion into a porous medium. Finally, they reach a well assumed to be a source of potable water for local residents. The annual individual dose is assessed on this well scenario to find out the significance of an EDZ. A profound sensitivity study was performed, but all results show that the impact is negligible. Even though the role of an EDZ turns out to be limited on overall safety assessment, still it is worthwhile to study the chemical role of an EDZ, such as a potential source for natural colloids, potential sealing of an open fracture by fine clay particles generated by the process of an EDZ, and alteration of a sorption mechanism by an EDZ in the future.