• 한국정보디스플레이학회:학술대회논문집
• /
• 2002.08a
• /
• pp.590-593
• /
• 2002

Using two- and three-dimensional fluid simulation codes, we have suggested several new plasma display panel (PDP) cell structures that have high luminous efficiency compared with conventional structure. To improve the luminance and discharge efficiency, we utilize long discharge path, lower electric field region, and reduction of power consumption by adding one auxiliary electrode or reducing the electrode area. Consequently, luminous efficiency increases about 1.8 times. Furthermore for the resonance radiation trapping effect in PDP system, we have described a self-consistent radiation transport model coupled with fluid simulation using modified Holstein's equation.

• International Journal of Computer Science & Network Security
• /
• v.21 no.2
• /
• pp.40-43
• /
• 2021

In the article of instability on the peak power level, duration and repetition period of a multifrequency space-time signal, we calculated the maximum values of the errors of the parameters of the laws of spatial-phase-frequency control. Requirements for the accuracy of the location of the phase centers of the emitters in a cylindrical phased array antenna with pyramidal horns; it is advisable to calculate the radiation field using single-stage and multi-stage distribution laws. The phase centers of individual radiation sources of a cylindrical phased array antenna have been studied; they have almost no effect on the duration and period of recurrence.

• Bulletin of the Korean Space Science Society
• /
• 2004.10b
• /
• pp.347-350
• /
• 2004

A space radiation analysis has been used to evaluate an ability of electronic equipment boxes or spacecrafts to endure various radiation effects, so it helps design thicknesses of structure and allocate components to meet the radiation requirements. A comparison study of space radiation dose analysis programs SPENVIS Sectoring Tool (SST) and SIGMA II is conducted through some structure cases, simple sphere shell, box and representative satellite configurations. The results and a discussion of comparison will be given. A general comparison will be shown for understanding those programs. The both programs use the same strategy, solid angle sectoring with ray-tracing method to produce an approximate dose at points in representative simple and complex models of spacecraft structures. Also the particle environment data corresponding to mission specification and radiation transport data are used as input data. But there are distinctions between them. The specification of geometry model and its input scheme, the assignment of dose point and the numbers, the prerequisite programs and ways of representing results will be discussed. SST is a web-based interactive program for sectoring analysis of complex geometries. It may be useful for a preliminary dose assessment with user-friendly interfaces and a package approach. SIGMA II is able to obtain from RSICC (Radiation Safety Information Computational Center) as a FOR-TRAN 77 source code. It may be suitable for either parametric preliminary design or detailed final design, e.g. a manned flight or radiation-sensitive component configuration design. It needs some debugs, recompiling and a tedious work to make geometrical quadric surfaces for actual spacecraft configuration, and has poor documentation. It is recommend to vist RSICC homepage and GEANT4/SSAT homepage.

• Journal of Radiation Protection and Research
• /
• v.47 no.3
• /
• pp.111-133
• /
• 2022

The exciting advancement related to the "modeling of digital human" in terms of a computational phantom for radiation dose calculations has to do with the latest hype related to deep learning. The advent of deep learning or artificial intelligence (AI) technology involving convolutional neural networks has brought an unprecedented level of innovation to the field of organ segmentation. In addition, graphics processing units (GPUs) are utilized as boosters for both real-time Monte Carlo simulations and AI-based image segmentation applications. These advancements provide the feasibility of creating three-dimensional (3D) geometric details of the human anatomy from tomographic imaging and performing Monte Carlo radiation transport simulations using increasingly fast and inexpensive computers. This review first introduces the history of three types of computational human phantoms: stylized medical internal radiation dosimetry (MIRD) phantoms, voxelized tomographic phantoms, and boundary representation (BREP) deformable phantoms. Then, the development of a person-specific phantom is demonstrated by introducing AI-based organ autosegmentation technology. Next, a new development in GPU-based Monte Carlo radiation dose calculations is introduced. Examples of applying computational phantoms and a new Monte Carlo code named ARCHER (Accelerated Radiation-transport Computations in Heterogeneous EnviRonments) to problems in radiation protection, imaging, and radiotherapy are presented from research projects performed by students at the Rensselaer Polytechnic Institute (RPI) and University of Science and Technology of China (USTC). Finally, this review discusses challenges and future research opportunities. We found that, owing to the latest computer hardware and AI technology, computational human body models are moving closer to real human anatomy structures for accurate radiation dose calculations.

• 한국ITS학회:학술대회논문집
• /
• 2004.11a
• /
• pp.216-219
• /
• 2004

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4280-4295
• /
• 2022

In this paper, a new multi-group neutron-gamma transport calculation code system STRAUM-MATXST for complicated geometrical problems is introduced and its development status including numerical tests is presented. In this code system, the MATXST (MATXS-based Cross Section Processor for S_N Transport) code generates multi-group neutron and gamma cross sections by processing MATXS format libraries generated using NJOY and the STRAUM (S_N Transport for Radiation Analysis with Unstructured Meshes) code performs multi-group neutron-gamma coupled transport calculation using tetrahedral meshes. In particular, this work presents the recent implementation and its test results of the Krylov subspace methods (i.e., Bi-CGSTAB and GMRES(m)) with preconditioners using DSA (Diffusion Synthetic Acceleration) and TSA (Transport Synthetic Acceleration). In addition, the Krylov subspace methods for accelerating the energy-group coupling iteration through thermal up-scatterings are implemented with new multi-group block DSA and TSA preconditioners in STRAUM.

• Journal of Radiation Protection and Research
• /
• v.28 no.4
• /
• pp.281-290
• /
• 2003

The KN-12 spent nuclear fuel transport cask, which is a Type B(U) package designed to comply with the requirements of Korea Atomic Energy Act[1], IAEA Safety Standards Series No.TS-R-1[2] and US 10 CFR Part 71[3], is designed for carrying up to 12 PWR spent fuel assemblies in a basket structure. The cask has been licensed in accordance with Korea Atomic Energy Act and was fabricated in Korea in accordance with the requirements of ASME B&PV Sec.III, Div.3[4]. The cask must maintain thermal integrity in accordance with the related regulations and be evaluated to verify that the thermal performance of the cask complies with the regulatory requirements. The temperatures of the cask and components were determined by using finite elements methods with a numerical tool, safety tests using an 1/8 height slice model of the real cask were conducted to demonstrate verification of the numerical tool and methods, and heat transfer tests for normal transport conditions were performed as a fabrication acceptance test to demonstrate the heat transfer capability of the cask.

• Journal of Mechanical Science and Technology
• /
• v.16 no.12
• /
• pp.1725-1732
• /
• 2002

Combined convection and radiation heat transfer in a circular tube with circumferential fins and circular disks is investigated for various operating conditions. Using a finite volume technique for steady laminar flow, the governing equations are solved in order to study the flow and temperature fields. The P- 1 approximation and the weighted sum of gray gases model (WSGGM) are used for solving the radiation transport equation. The results show that the total Nusselt number of combined convection and radiation is higher than that of pure convection. If the temperatures of the combustion gas and the wall in a tube are high, radiation becomes dominant. Therefore, it is necessary to evaluate the effect of radiation on the total heat transfer.

• Nuclear Engineering and Technology
• /
• v.54 no.4
• /
• pp.1449-1455
• /
• 2022

A radiotracer investigation was carried out in a Thermal Stratification Test Facility (TSTF) with objectives of investigating the dispersion and diffusion of radionuclide and effectiveness of the thermocline to minimize the radionuclide content in the hot water layer. Technetium-99m (^99mTc) as sodium pertechnetate was used as a radiotracer in the investigation. Qualitative analysis showed that a thermocline is formed within the TSTF and is effective in preventing the transport of radionuclide from bottom section to the top section of the facility. It was found that the radiotracer injected at the bottom of the pool took about 17.4 h to disperse from bottom to the top of the facility. The results of the investigation helped in understanding the effectiveness of hot water layer and thus to minimize the pool top radiation levels.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.4137-4141
• /
• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.