• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Quality Improvement in Health Care
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• v.20 no.1
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• pp.60-73
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• 2014

Objectives: This study aims at decreasing spatial dose rate through work improvement whilst spatial dose rate is the cause of increasing personal exposure dose which occurs in the process of handling radioisotope. Methods: From February 2013 until July 2013, divided into "before" and "after" the improvement, spatial dose rate in laboratory of nuclear medicine was measured in gamma image room, PET/CT-1 image room, and PET/CT-2 image room as its locations. The measurement time was 08:00, 12:00 and 17:00, and SPSS 21.0 USA was opted for its statistical analysis. Result: The spatial dose rate at distribution worktable, injection table, the entrance to the distribution room, and radioisotope storage box, which had showed high spatial dose rate, decreased by more than 43.7% a monthly average. The distribution worktable, that had showed the highest spatial dose rate in PET/CT-1 image room, dropped the rate to 42.3% as of July. The injection table and distribution worktable in the PET/CT-2 image room also showed the decline of spatial dose rate to 89% and 64.4%, respectively. Conclusion: By improving distribution process and introducing proper radiation shielding material, we were able to drop the spatial dose rate substantially at distribution worktable, injection table, and nuclide storage box. However, taking into account of steadily increasing amount of radioisotope used, strengthening radiation related regulations, and safe utilization of radioisotope, the process of system improvement needs to be maintained through continuous monitoring.

• Journal of Radiation Protection and Research
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• v.15 no.2
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• pp.7-16
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• 1990

The sorption of radioactive cobalt and ruthenium on alumina, silica gel, zeolite 3A, kaolin and Na-bentonite has been studied as a function of pH. nuclide concentration and ionic strength. Retardation factor for cobalt and ruthenium on soil minerals was determined through porosity measurement. Hydrolysed species, cobalt and ruthenium interact with solid surfaces by physical adsorption processes. Freundlich sorption isotherms for cobalt and ruthenium are effectively linear. The sorption decreases with increasing ionic strength for cobalt and ruthenium. The effect of increasing porosity on the retardation factor countered the effect of a significant increase in the distribution coefficient.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.303-309
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• 2019

Tritium is an important nuclide that must be monitored for radiation safety management. In this study, HTO was orally administered to rats at the level of 37 kBq ($1{\mu}Ci$) or 370 kBq ($10{\mu}Ci$) to examine tissue distribution and excretion levels. After sacrifice, wet and dry tissue samples were weighed and analyzed for tissue free-water tritium (TFWT) and organically bound tritium (OBT). The mean tissue concentrations of TFWT (OBT) were 30.9 (17.8) and 4.4 (8.1) Bq/g on days 7 and 13 at the 37 kBq level and 30.8 (64.6) Bq/g on day 17 at the 370 kBq level. To assess the cytogenetic damage due to tritium exposure, a cytokinesis-blocked micronucleus (MN) assay was performed in blood samples from rats exposed to HTO for 14 and 21 days after oral administration. There was no significant difference in the MN frequencies between the control and exposed rats.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.32-37
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• 2005

The characteristics of the metal melting and radionuclide distribution of the radioactive has been investigated in a lab-scale arc furnace. The slag former based on the constituents of silica, calcium oxide, aluminum oxide, borate and calcium fluoride additions was used for melting of the stainless and carbon steel. In the melting of the stainless steel, the amount of slag formation increased with an increase of the concentration of the slag former. But the effects of the slag basicity on the amount of stag formation showed a local maximum value of the slag formation with an increase of the basicity index in the melting of the stainless steel as well as in the melting of the carbon steel. With an increase of the amount of slag former addition, the trends of the cobalt distribution into the ingot and the stag depended on the kind of slag former used in the melting of the stainless steel while the effect of the slag basicity on the distribution of the cobalt was not clarified in the melting of carbon steel. Tn the melting of the carbon steel, the strontium was captured at up to $50\%$ into the slag phase. Cesium was completely eliminated from the melt of the stainless steel as well as the carbon steel and distributed to the dust phase.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.179-194
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• 2020

A Bayesian approach was introduced to improve the belief of prior distributions of input parameters for the probabilistic safety assessment of radioactive waste repository. A GoldSim-based module was developed using the Markov chain Monte Carlo algorithm and implemented through GSTSPA (GoldSim Total System Performance Assessment), a GoldSim template for generic/site-specific safety assessment of the radioactive repository system. In this study, sequential Bayesian updating of prior distributions was comprehensively explained and used as a basis to conduct a reliable safety assessment of the repository. The prior distribution to three sequential posterior distributions for several selected parameters associated with nuclide transport in the fractured rock medium was updated with assumed likelihood functions. The process was demonstrated through a probabilistic safety assessment of the conceptual repository for illustrative purposes. Through this study, it was shown that insufficient observed data could enhance the belief of prior distributions for input parameter values commonly available, which are usually uncertain. This is particularly applicable for nuclide behavior in and around the repository system, which typically exhibited a long time span and wide modeling domain.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2012.05a
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• pp.266-269
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• 2012

Over the world, the development of critical technology for the safety and security of coastal distribution based on ubiquitous computing is being processed to meet the bottom lines of governmental regulations. Currently, 3D detectors using X-rays for freight search technology are being developed concerning U-based shipping security systems. In addition to that, there are increasing needs for developing portable radiation detector to detect gamma rays nuclide. The purpose of the thesis is to suggest the research of high speed portable gamma rays detector platform to prevent disorder caused to humans and public objects.

• Journal of radiological science and technology
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• v.45 no.5
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• pp.433-438
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• 2022

Targeted radionuclides treatment (TRT) requires the establishment of treatment plans that consider various factors, such as the type of radionuclides, target organs, and administration methods. For this reason, in this study, the absorption dose of a single cell was analyzed according to the type of radioisotope used to treat target radionuclides. In this study, a simulation was performed on beta rays used in the treatment of target radionuclides at the cell level using MCNPX (ver. 2.5.0). First, according to the calculation formula, the beam path according to the type of radioisotope for treatment was calculated. Second, the amount of self-radiation by beta rays emitted from cell diameters of 5 ㎛ and 10 ㎛ cell nuclei was evaluated. As a result, it showed a high range proportional to the maximum energy of the beta-ray, and the highest self-dose distribution from 177 Lu radiation sources among therapeutic radioisotopes. This was analyzed as a result that is inversely proportional to the maximum energy of the beta-ray, and it suggests that the selection of a nuclide considering the range of the beta-ray is necessary in the treatment of target radionuclides in the future.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3888-3896
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• 2022

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm k_eff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of ⁵²Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

• Journal of Radiation Protection and Research
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• v.46 no.2
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• pp.58-65
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• 2021

Background: The evaluation of skyshine distribution, release of airborne radioactive nuclides, and soil activation and groundwater migration were required for radiological assessment of the impact on the environment surrounding In-Flight (IF)-system facility of the RAON (Rare isotope Accelerator complex for ON-line experiment) accelerator complex. Materials and Methods: Monte Carlo simulation by MCNPX code was used for evaluation of skyshine and activation analysis for air and soil. The concentration model was applied in the estimation of the groundwater migration of radionuclides in soil. Results and Discussion: The skyshine dose rates at 1 km from the facility were evaluated as 1.62 × 10^-3 μSv·hr^-1. The annual releases of ³H and ¹⁴C were calculated as 9.62 × 10^-5 mg and 1.19 × 10^-1 mg, respectively. The concentrations of ³H and ²²Na in drinking water were estimated as 1.22 × 10^-1 Bq·cm^-3 and 8.25 × 10^-3 Bq·cm^-3, respectively. Conclusion: Radiological assessment of environmental impact on the IF-facility of RAON was performed through evaluation of skyshine dose distribution, evaluation of annual emission of long-lived radionuclides in the air and estimation of soil activation and groundwater migration of radionuclides. As a result, much lower exposure than the limit value for the public, 1 mSv·yr^-1, is expected during operation of the IF-facility.