• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.331-341
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• 2016

To evaluate the physicochemical and radiological properties of raw materials and by-products in domestic distribution, about 220 samples with 16 species were prepared. We measured the energy spectrum and the chemical content, such as U, Th, and K, using a $LaBr_3$ scintillation detector and ED-XRF. In addition, HPGe detector was used to analyze the radioac-tivity of $^{234}Th$, $^{234}mPa$, and $^{214}Bi$ in uranium decay series and $^{228}Ac$, $^{212}Pb$, and $^{208}Tl$ in thorium decay series, and $^{40}K$. The correlation between characteristic variables, such as the count rate in several ROIs, chemical content, and radioactivity, was assessed to infer the radioactivity of natural radionuclides through a rapid screening method. Based on the results, a characteristic database for raw material and by-product in domestic distribution was established and it will provide useful information in the analysis procedure and improve the accuracy and reproducibility in the analysis of natural radionuclides.

• Journal of the Korean Society of Radiology
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• v.15 no.3
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• pp.391-399
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• 2021

Eu²⁺ doped polyphosphate NaSr(PO₃)₃ blue-emitting phosphors were synthesized by the conventional solid state method in a reductive atmosphere. The phase formation of NaSr(PO₃)₃ phosphors were characterized by using the X-ray powder diffraction (XRD) measurement. The photoluminescence emission and excitation spectra of the NaSr(PO₃)₃:Eu²⁺ phosphor, and decay curves were measured. Under the near-UV excitation, the phosphor exhibits a band emission around 420 nm assigned to the 4f⁶5d→f7(8S^7/2) transition of Eu²⁺. The temperature dependent emission spectra and decay curves were measured to elevate the thermal properties of the Eu²⁺ doped phosphors. The as-prepared NaSr(PO₃)₃:Eu²⁺ phosphors show a strong temperature dependent performance, which can serve as a promising temperature sensor.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1540-1554
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• 2023

To use nuclear energy sustainably, spent nuclear fuel, classified as high-level radioactive waste and inevitably discharged after electricity generation by nuclear power plants, must be managed safely and isolated from the human environment. In Korea, the land area is limited and the amount of high-level radioactive waste, including spent nuclear fuels to be disposed, is relatively large. Thus, it is particularly necessary to maximize disposal efficiency. In this study, a high-efficiency deep geological repository concept was developed to enhance disposal efficiency. To this end, design strategies and requirements for a high-efficiency deep geological repository system were established, and engineered barrier modules with a disposal canister for pressurized water reactor (PWR)-type and pressurized heavy water reactor type Canada deuterium uranium (CANDU) plants were developed. Thermal and structural stability assessments were conducted for the repository system; it was confirmed that the system was suitable for the established strategies and requirements. In addition, the results of the nuclear safety assessment showed that the radiological safety of the new system met the Korean safety standards for disposal of high-level radioactive waste in terms of radiological dose. To evaluate disposal efficiency in terms of the disposal area, the layout of the developed disposal areas was assessed in terms of thermal limits. The estimated disposal areas were 2.51 km² and 1.82 km² (existing repository system: 4.57 km²) and the excavated host rock volumes were 2.7 Mm³ and 2.0 Mm³ (existing repository system: 4.5 Mm³) for thermal limits of 100 ℃ and 130 ℃, respectively. These results indicated that the area and the excavated volume of the new repository system were reduced by 40-60% compared to the existing repository system. In addition, methods to further improve the efficiency were derived for the disposal area for deep geological disposal of spent nuclear fuel. The results of this study are expected to be useful in establishing a national high-level radioactive waste management policy, and for the design of a commercial deep geological repository system for spent nuclear fuels.

• Progress in Medical Physics
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• v.25 no.1
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• pp.31-36
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• 2014

We aimed to evaluate the dosimetric characteristics of reproducibility, linearity and dose dependence of optical stimulated luminance dosimeter (OSLD) in the Co-60 Gamma-rays and to analyze with a precedent study in field of the diagnostic radiography and radiotherapy. The reproducibility was 0.76% of the coefficient of variation, the homogeneity was within 1.5% of the coefficient of variation and OSLD had supra-linear response more than 3 Gy. So the correlation between dose and count was fitted by quadratic function. The count depletion by repeated reading was 0.04% per reading regardless of the irradiated dose. And the half time of decay curve according to the irradiated dose was 0.68 min. with 1 Gy, 1.04 min. with 5 Gy, and 1.10 min. with 10 Gy, respectively. In case of annealing for 30 min, the removal rate was 88% with 1 Gy, 90% with 5 Gy, and 92% with 10 Gy, respectively and 99% in case of annealing time for 4 hour. It is feasible to use OSLDs for dose evaluation in Co-60 Gamma-rays when considering the uncertainty on the procedure according to the irradiated dose.

• Investigative Magnetic Resonance Imaging
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• v.20 no.4
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• pp.207-214
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• 2016

Purpose: The purpose of this study was to investigate the effect of the number of measurement points on the calculation of transverse relaxation time (T2) with a focus on muscle T2. Materials and Methods: This study assumed that muscle T2 was comprised of a single component. Two phantom types were measured, 1 each for long ("phantom") and short T2 ("polyvinyl alcohol gel"). Right calf muscle T2 measurements were conducted in 9 healthy male volunteers using multiple-spin-echo magnetic resonance imaging. For phantoms and muscle (medial gastrocnemius), 5 regions of interests were selected. All region of interest values were expressed as the mean ${\pm}$ standard deviation. The T2 effective signal-ratio characteristics were used as an index to evaluate the magnetic resonance image quality for the calculation of T2 from T2-weighted images. The T2 accuracy was evaluated to determine the T2 reproducibility and the goodness-of-fit from the probability Q. Results: For the phantom and polyvinyl alcohol gel, the standard deviation of the magnetic resonance image signal at each echo time was narrow and mono-exponential, which caused large variations in the muscle T2 decay curves. The T2 effective signal-ratio change varied with T2, with the greatest decreases apparent for a short T2. There were no significant differences in T2 reproducibility when > 3 measurement points were used. There were no significant differences in goodness-of-fit when > 6 measurement points were used. Although the measurement point evaluations were stable when > 3 measurement points were used, calculation of T2 using 4 measurement points had the highest accuracy according to the goodness-of-fit. Even if the number of measurement points was increased, there was little improvement in the probability Q. Conclusion: Four measurement points gave excellent reproducibility and goodness-of-fit when muscle T2 was considered mono-exponential.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.165-169
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• 2018

An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was $7.40{\times}10^5Bq/cm^3$ and at the bay region was $3.39{\times}10^5Bq/cm^3$. At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was $4.14{\times}10^{-1}Bq/cm^3$.

• Journal of Sensor Science and Technology
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• v.19 no.6
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• pp.443-448
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• 2010

In this work, the scintillation properties of CsI:Na crystal were investigated as radiation detection sensor. This scintillation material was grown by a 2-zone vertical Bridgman method. Under X-ray excitation the crystal shows a broad emission band between 280 nm and 690 nm wavelength range, peaking at 413 nm. Energy resolution for $^{137}Cs$ 662 keV $\gamma$-rays of the crystal was measured to be 6.9 %(FWHM). At room temperature, the crystal exhibits three exponential decay time components. The fast and major component of scintillation time profile of the crystal emission decays with a 457 ns time constant. Absolute light yield of the crystal was estimated to be 53,000 ph/MeV using LAAPD. The sample crystal shows proportionality of 30 % in the measured energy range from 31 to 1,333 keV. And the $\alpha/\beta$ ratio of the crystal was 0.14.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.1-8
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• 2010

Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. These conceptual uncertainties are additional to those that more typically arise, for example, in the assessment of present day situations, but which also have to be addressed. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2771-2782
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• 2022

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.