• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.364-364
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• 2002

• Progress in Medical Physics
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• v.10 no.1
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• pp.1-7
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• 1999

In this study, the dose distributions of a $^{32}$ p uniform cylindrical volume source and a surface source, a pure $\beta$emitter, were calculated in order to obtain information relevant to the utilization of a balloon catheter and a radioactive stent. The dose distributions of $^{32}$ p were calculated by means of the EGS4 code system. The sources are considered to be distributed uniformly in the volume and on the surface in the form of a cylinder with a radius of 1.5 mm and length of 20 mm. The energy of $\beta$particles emitted is chosen at random in the $\beta$ energy spectrum evaluated by the solution of the Dirac equation for the Coulomb potential. Liquid water is used to simulate the particle transport in the human body. The dose rates in a target at a 0.5mm radial distance from the surface of cylindrical volume and surface source are 12.133 cGy/s per GBq (0.449 cGy/s per mCi, uncertainty: 1.51%) and 24.732 cGy/s per GBq (0.915 cGy/s per mCi, uncertainty: 1.01%), respectively. The dose rates in the two sources decrease with distance in both radial and axial direction. On the basis of the above results, the determined initial activities were 29.69 mCi and 1.2278 $\mu$Ci for the balloon catheter and the radioactive stent using $^{32}$ P isotope, respectively. The total absorbed dose for optimal therapeutic regimen is considered to be 20 Gy and the treatment time in the case of the balloon catheter is less than 3 min. Absorbed doses in targets placed in a radial direction for the two sources were also calculated when it expressed initial activity in a 1 mCi/ml volume activity density for the cylindrical volume source and a 0.1 mCi/cm$^2$ area activity density for the surface source. The absorbed dose distribution around the $^{32}$ P cylindrical source with different size can be easily calculated using our results when the volume activity density and area activity density for the source are known.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.590-595
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• 2003

An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for a PWR spent fuel assembly. The axial background neutron flux is measured in a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of the Poisson regression to the net induced fission neutron counts. The measured keffs determined on the basis of the exponential decay constants of Cl5 appeared to be 0.541, 0.540, 0.597 and 0.556, respectively, which are comparable with 0.55195$\pm$0.00232 of the MCNP calculation.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.315-325
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• 1990

The analytical repository source term (REPS) computer code is developed for the safety assessment of radioactive waste geologic repository. For reliable prediction of the leach rates for various radionuclides, degradation of concrete structures, corrosion rate of waste container, degree of corrosion on the container surface, and the characteristics of radionuclides are considered in this REPS code. For the validation of the radionuclide leach rates predicted by the REPS model, the calculated leach rates of Cs-137, Sr-85, and Co-60 are compared with two reported leaching test results. Cesium and strontium leach congruently, and the leaching test results of these species can be reproduced by the congruent leaching model included in the REPS model. In case of cobalt, the solid diffusion model is in good agreement with the leaching test results.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.76-85
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• 2005

The conceptual study of Accelerator Driven System (ADS) had lasted for about five years and ended in 1999 in China. As one project of 'the major state basic research program (973)' in energy domain, which is sponsored by the China Ministry of Science and Technology (MOST), a five years program of basic research for ADS physics and related technology has been launched since 2000 and passed national review last month. CIAE (China Institute of Atomic Energy), IHEP (Institute of High Energy Physics), PKU-IHIP (Institute of Heavy Ion Physics in Peking University) and other institutions are jointly carrying on the research. The research activities are focused on HPPA physics and technology, reactor physics of external source driven sub-critical assembly, nuclear data base and material study. For HPPA, a high current injector consisting of an ECR ion source, LEBT and a RFQ accelerating structure of 3.5MeV has been built. In reactor physics study, a series of neutron multiplication experimental study has been carried out and is being carrying on. The VENUS facility has been constructed as the basic experimental platform for the neutronics study in ADS blanket. It's a zero power sub-critical neutron multiplying assembly driven by external neutron produced by a pulsed neutron generator. The theoretical, experimental and simulation study on nuclear data, material properties and nuclear fuel circulation related to ADS is carrying on to provide the database for ADS system analysis. The main results on ADS related researches will be reported.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2734-2741
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• 2023

Brachytherapy is a special case of radiotherapy. It should be arranged according to some principles in medical radiation applications and radiation physics. The primary principle is to use as low as reasonably achievable dose in all ionizing radiation applications for diagnostic and therapeutic treatments. Dosimetric distributions are dependent on radioactive source properties and radiation-matter interactions in an absorber medium such as phantom or tissue. In this consideration, the geometrical structure and material of the seed capsule, which surrounds a radioactive material, are directly responsible for isodose profiles and dosimetric functions. In this study, the radiometric properties of capsule material were investigated on dose distribution in a water phantom by changing its nuclear properties using the EGSnrc Monte Carlo (MC) simulation code. Effective atomic numbers of hypothetic mixtures were calculated by using different elements with several fractions for capsule material. Model 6711 brachytherapy seed was modeled by EGSnrc/Dosrcnrc Code and dosimetric functions were calculated. As a result, dosimetric parameters of hypothetic sources have been acquired in large-scale atomic number. Dosimetric deviations between the data of hypothetic seeds and the original one were analyzed. Unit dose (Gy/Particle) distributions belonging to different types of material in seed capsule have remarkably differed from the original capsule's data. Capsule type is major variable to manage the expected dose profile and isodose distribution around a seed. This study shows us systematically varied scale of material type (cross section or effective atomic number dependent) offers selective material usage in production of seed capsules for the expected isodose profile of a specific source.

• 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.

• Journal of the Korean Society of Radiology
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• v.15 no.6
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• pp.813-819
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• 2021

In this study, the shielding rate of L-block-type shielding equipment used for radiation protection when radioactive fluorine is injected into the human body and the dose distribution of the space in the injection room were calculated using the Monte Carlo method. The shielding rate of the body and window parts of the L-block-type shielding equipment was 99.99%. The dose distribution calculated at a distance of 1 m was relatively high at 135°, 45°, 225°, 315°, and 180° of the XZ plane, and was calculated to be very low at 0°, 90°, and 270°. In the YZ plane, it was relatively high at 135°, 180°, and 225°, and was calculated very low at the remaining angles. The AZ and BZ planes also showed similar results to the YZ plane. In addition, it was confirmed that the shielding rate was the best in the range of 225° to 315° through the dose distribution in the horizontal direction of the source and the 45° direction above the source. These results can be used as basic data necessary for radiation protection of radiation workers.

• Geomechanics and Engineering
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• v.18 no.1
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• pp.11-20
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• 2019

The purpose of this study was to propose a new approach for quantifying in situ rock mass damage, which would include a degree-of-damage and the degraded strength of a rock mass, along with its prediction based on real-time Acoustic Emission (AE) observations. The basic approach for quantifying in-situ rock mass damage is to derive the normalized value of measured AE energy with the maximum AE energy, called the degree-of-damage in this study. With regard to estimation of the AE energy, an AE crack source location algorithm of the Wigner-Ville Distribution combined with Biot's wave dispersion model, was applied for more reliable AE crack source localization in a rock mass. In situ AE wave attenuation was also taken into account for AE energy correction in accordance with the propagation distance of an AE wave. To infer the maximum AE energy, fractal theory was used for scale-independent AE energy estimation. In addition, the Weibull model was also applied to determine statistically the AE crack size under a jointed rock mass. Subsequently, the proposed methodology was calibrated using an in situ test carried out in the Underground Research Tunnel at the Korea Atomic Energy Research Institute. This was done under a condition of controlled incremental cyclic loading, which had been performed as part of a preceding study. It was found that the inferred degree-of-damage agreed quite well with the results from the in situ test. The methodology proposed in this study can be regarded as a reasonable approach for quantifying rock mass damage.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.281-291
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• 2013

A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.