• Radiation Oncology Journal
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• v.21 no.2
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• pp.158-166
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• 2003

Purpose: To propose a conceptual design of a novel source for intensity modulated brachytherapy. Materials and Methods: The source design incorporates both radioactive and shielding materials (stainless steel or tungsten), to provide an asymmetric dose intensity in the azimuthal direction. The intensity modulated intravascular brachytherapy was performed by combining a series of dwell positions and times, distributed along the azimuthal coordinates. Two simple designs for the beta-emitting sources, with similar physical dimensions to a $^{90}Sr/Y$ Novoste Beat-Cath source, were considered in the dosimetric feasibility study. In the first design, the radioactive and materials each occupy half of the cylinder and in the second, the radioactive material occupies only a quater of the cylinder. The radial and azimuthal dose distributions around each source were calculated using the MCNP Monte Carlo code. Results: The preliminary hypothetical simulation and optimization results demonstrated the 87$\%$ difference between the maximum and minimum doses to the lumen wall, due to off-centering of the radiation source, could be reduced to less than 7$\%$ by optimizing the azimuthal dwell positions and times of the partially shielded intravascular brachytherapy sources. Conclusion: The novel brachytherapy source design, and conceptual source delivery system, proposed in this study show promising dosimetric characteristics for the realization of intensity modulated brachytherapy in intravascular treatment. Further development of this concept will center on building a delivery system that can precisely control the angular motion of a radiation source in a small-diameter catheter.

• Journal of Radiation Protection and Research
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• v.29 no.3
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• pp.157-163
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• 2004

Atmospheric dispersion modelling has been widely used to predict the fate and transport of radioactive or toxic materials released from nuclear facilities which is an unlikely accidental event. To improve the forecasting performance of the dispersion model, it is required that source rate and dispersion characteristics must be defined appropriately. Generally, source term of the radioactive materials is much uncertain at the early phase of an accidental event. In this study, we computed the source rate with the experimental field data monitored at the Yeoung-Kwang nuclear site and obtained the optimal source rate to minimize the errors between the measured concentrations and the computed ones by the Gaussian plume model. Computed source term showed a good result within 24% of the artificially released source rate.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.833-838
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• 1995

This study presents a method to predict plant-specific and operational history dependent scaling factors. Realistic and detailed approaches are taken to find scaling factors at reactor coolant. This approach begins with fission product release mechanisms and fundamental release properties of fuel-source nuclide such as fission product and transuranic nuclide. Scaling factors at various waste streams are derived from the predicted reactor coolant scaling factors with the use of radionuclide retention and build up model. This model makes use of radioactive material balance within the radioactive waste processing systems. According to input parameters of plant operation history, scaling factors predicted at reactor coolant and waste streams are well brought out the effects of plant operation history.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.232-241
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• 2002

A safety assessment is carried out for the near-surface radioactive waste disposal in the reference engineered vault facility. The analysis is mainly divided into two parts. One deals with the release and transport of radionuclide in the vault and unsaturated zone. The other deals with the transport of radionuclide in the saturated zone and radiological impacts to a human group under well drinking water scenario. The parameters for source-term, geosphere and biosphere models are mainly obtained from the site specific data. The results show that the annual effective doses are dominated by long lived, mobile radionuclides and their associated daughters. And it is found that the total effective dose for drinking water is far below the general criteria of regulatory limit for radioactive waste disposal facility.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.310-314
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• 2016

Background: FCTC10 container is designed to transport $^{60}Co$ radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a $^{60}Co$ source of $1.8{\times}10^5$ Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1966-1973
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• 2023

Virtual reality technology has been widely used in the field of nuclear and radiation safety, dose rate computing in virtual environment is essential for optimizing radiation protection and planning the work in radioactive-controlled area. Because the CPU-based gamma dose rate computing takes up a large amount of time and computing power for voxelization of volumetric radioactive source, it is inefficient and limited in its applied scope. This study is to develop an efficient gamma dose rate computing code and apply into fast virtual simulation. To improve the computing efficiency of the point kernel algorithm in the reference (Li et al., 2020), we design a GPU-based computing framework for taking full advantage of computing power of virtual engine, propose a novel voxelization algorithm of volumetric radioactive source. According to the framework, we develop the GPPK(GPU-based point kernel gamma dose rate computing) code using GPU programming, to realize the fast dose rate computing in virtual world. The test results show that the GPPK code is play and plug for different scenarios of virtual simulation, has a better performance than CPU-based gamma dose rate computing code, especially on the voxelization of three-dimensional (3D) model. The accuracy of dose rates from the proposed method is in the acceptable range.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4204-4212
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• 2023

Background: Storage reservoirs of radioactive waste could be the source of atmospheric pollution due to the efflux of aqueous aerosol from their water areas. The main mechanism of formation of aqueous aerosols is the collapse of gas bubbles at the water surface. In this paper, we discuss the potential influence of biological factors on gas ebullition in the water areas of the radioactively contaminated industrial reservoirs R-9 (Lake Karachay) and R-4 (Metlinsky pond) of the Mayak PA. The emission of the released non-dissolved gases captured with gas traps in reservoir R-9 was (88-290) ml/m² per day (2015) and in reservoir R-4 (270-460) ml/m² per day (2016). The analysis of gas composition in reservoir R-4 (60% methane, 35% nitrogen, 2.4% oxygen, 1.5% carbon dioxide) confirms their biological origin. It is associated with the processes of organic matter destruction in bottom sediments. The major source of organic matter in bottom sediments is the dying phytoplankton developing in these reservoirs. Conclusion: The obtained results form the basis to set a task to quantify the relationship between the phytoplankton development, gases ebullition and radioactive atmosphere contamination.

• Journal of Radiation Protection and Research
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• v.45 no.3
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• pp.117-129
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• 2020

Background: Radiation portal monitors (RPMs) involving plastic scintillators installed at the border inspection sites can detect illicit trafficking of radioactive sources in cargo containers within seconds. However, RPMs may generate false alarms because of the naturally occurring radioactive materials. To manage these false alarms, we previously suggested an energy-weighted algorithm that emphasizes the Compton-edge area as an outstanding peak. This study intends to evaluate the identification of radioactive sources using an improved energy-weighted algorithm. Materials and Methods: The algorithm was modified by increasing the energy weighting factor, and different peak combinations of the energy-weighted spectra were tested for source identification. A commercialized RPM system was used to measure the energy-weighted spectra. The RPM comprised two large plastic scintillators with dimensions of 174 × 29 × 7 ㎤ facing each other at a distance of 4.6 m. In addition, the in-house-fabricated signal processing boards were connected to collect the signal converted into a spectrum. Further, the spectra from eight radioactive sources, including special nuclear materials (SNMs), which were set in motion using a linear motion system (LMS) and a cargo truck, were estimated to identify the source identification rate. Results and Discussion: Each energy-weighted spectrum exhibited a specific peak location, although high statistical fluctuation errors could be observed in the spectrum with the increasing source speed. In particular, ¹³⁷Cs and ⁶⁰Co in motion were identified completely (100%) at speeds of 5 and 10 km/hr. Further, SNMs, which trigger the RPM alarm, were identified approximately 80% of the time at both the aforementioned speeds. Conclusion: Using the modified energy-weighted algorithm, several characteristics of the energy weighted spectra could be observed when the used sources were in motion and when the geometric efficiency was low. In particular, the discrimination between ⁶⁰Co and ⁴⁰K, which triggers false alarms at the primary inspection sites, can be improved using the proposed algorithm.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.73-80
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• 2011

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source ($^{60}Co$ radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.