• Radiation Oncology Journal
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• v.17 no.1
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• pp.23-29
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• 1999

Purpose : To evaluate the efficacy of radiation therapy and extrafascial hysterectomy in bulky stage IB, IIA-B uterine cervix cancers. Methods and Materials : Twenty four patients with bulky stage IB and IIA-B carcinoma of the uterine cervix were treated with extrafascial hysterectomy following radiation therapy due to doubts of residual disease at Department of therapeutic radiology, Keimyung University, Dongsan Hospital, from April 1986 to December 1997 According to FIGO staging system, there were 7 patients with stage IB, 9 patients with IIA and 8 patients with IIB stage whose median age was 45. Pathologic distribution showed 16 patients with squamous cell carcinoma and 8 patients with adenocarcinoma. Seven patients had tumors that are less than 5cm in size and 17 patients had tumors with larger than 5cm. The mean interval between radiation therapy and extrafascial hysterectomy was 57 days. The radiation therapy consisted of external irradition to the whole pelvis (180 cGy/fraction, mean 4100 cGy) and parametrial boost (for a mean total dose of 5000 cGy) with midline shield (4H 10 cm), followed by intracavitary irradiation up to 7500 cGy to point A (maximum 8500 cGy). The maximum follow up duration was 107 months and mean follow up duration was 42 months. Results :Ten out of 24 patients (41.7%) had residual disease found at the time of extrafascial hysterectomies. Five year overall survival rate (5Y OSR) and five year disease free survival rate (5Y DFSR) were 63.6% and 62.5% respectively. Five year overall survival rate for stage IB and IIA was 71.4% and 50% for stage IIB. There was a significant difference in 5Y OSR and 5Y DFSR between patients with and those without residual disease (negative vs positive, 83.3% vs. 40% (P=0.01), 83.3% vs 36% (P=0.01) respectively). There was a notable tendency of better survival with adenocarcinoma than with squamous cell carcinoma (adenocarcinoma vs squamous cell carcinoma, 85.7% vs. 53.3% (P=0.1), 85.7% vs. 50.9% (P=0.1) of 5Y OSR and 5Y DFS respectivey). Total dose to A point did not make a significant difference in survival rate or the existence of residual lesion (< 7500 cGy, ${\geq}$ 7500 cOy). It was also noted that significantly more frequent local failures have occurred in patients with positive residual disease compared with negative residual disease (5/10 vs. 0/14, p=0.003), There was no death related to the treatment. Conclusion : There was no improvement of residual disease and to the overall survival rate in spite of increased total dose to point A. We conclude that there is a possible beneficial effect of radiation therapy followed by extrafaseial hysterectomy in survival for adenocarcinoma of bulky stage IB and IIA-B uterine cervix. We need to confirm this with longer follow up and with large number of patients.

• Journal of the Korean Society for Precision Engineering
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• v.33 no.4
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• pp.271-277
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• 2016

For a stable electric power supply in the space, nuclear batteries have been used as the main power source in a spacecraft owing to their long lifetime and high reliability. In accordance with the plan for lunar mission in Korea, nuclear batteries will supply electricity to the rover that needs to be developed. According to the information about the estimated payload, Korea Atomic Energy Research Institute started with the conceptual design based on the previous studies in USA and Russia. Because a nuclear battery converts the decay heat of the radioisotope into electricity, thermal design, radiation shield, and shock protection need to be considered. In this study, two types of nuclear batteries, radial type and axial type, were designed according to the alignment of the thermoelectric module. Heat transfer analyses were performed to compare their thermoelectric efficiency, and test mockups were fabricated to evaluate their performances.

• The Journal of the Korea Contents Association
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• v.16 no.10
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• pp.758-765
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• 2016

This study analyzed for the radioactive shielding wall, which shields the medical linear accelerator. This allows to evaluate the level of waste with respect to the shield wall, which accounts for more than half of the cost of dismantling later linac facility. In addition, by analyzing the waste processing method according we discuss the way to obtain the benefits in terms of dismantling cost. Results of the simulate, the amount sufficient to screen the amount of neutron radiation occurring in the shielding wall linac was measured. And neutron activation analysis results were analyzed nuclides more than about 20. This analysis was in excess of that, $^{24}Na$, $^{45}Ca$, $^{59}Fe$ nucleus paper deregulation concentration. The value is reduced is greater the deeper the depth of the shielding wall concentration. Based on this, three specific areas (E, F, G) was estimated to be impossible to landfill or recycling. The rest area was estimated to be buried or recycled if possible more than a predetermined depth.

• Journal of Hydrogen and New Energy
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• v.7 no.2
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• pp.121-128
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• 1996

A direct cooling equipment for hydrogen liquefaction has been developed and tested. A direct cooling equipment consists of a liquefaction vessel, a radiation shield, a cryostat and a GM refrigerator. The cool-down and warm-up characteristics of the liquefaction apparatus have been investigated in detail. It is found that the hydrogen starts to be liquefied in the liquefaction vessel after 45 minutes of cool-down. The cool-down and warm-up tests of helium gas are also performed. The cool-down and warm-up characteristics of helium gas are found to be very different from those of hydrogen gas, since helium is not liquefied under the present operating conditions. When the liquefaction vessel is evacuated, natural convection phenomena of charged gas in liquefaction vessel can be removed. It is seen that the cool-down time of liquefaction vessel is substantially increased in vacuum environment.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.584-589
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• 2003

The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.196-203
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• 1982

To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail car. The cask requires 33cm thick steel shield and 27cm thick water region to satisfy the 3 feet apart dose rate limit set forth in 10 CFR 71, and 1.27cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel cladding and centerline temperatures were calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. k$_{eff}$ was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radiation source terms were calculated using ORIGEN-79.9.

• Progress in Medical Physics
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• v.15 no.4
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• pp.247-254
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• 2004

According to developing high energy linear accelerators and treatment methods, like (3 dimensional conformal radiotherapy (3D-CRT), intensity modulated radiotherapy (IMRT), many radiotherapy centers are replacing older linear accelerators with new higher technical machines. This often presents a shielding problem as the designed shield for the existing rooms is not adequate for the higher technical machines. Additional shielding in limited existing space becomes necessary. We are replacing older brachytherapy room with new higher technical linear accelerator for IMRT. This room is not adequate for the IMRT machine without additional shielding design. The logical development of optimum structural shielding designs with concrete and high density shielding blocks are presented. We obtained following results by comparison between the pre-calculating values and actual survey of completed LINAC installation. High density shielding blocks have more powerful radiation protection about 2 times.

• Journal of the Korea Institute of Building Construction
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• v.17 no.1
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• pp.9-14
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• 2017

Mechanical properties and neutron shielding performance of concrete with amorphous boron steel fiber have been investigated in this study. The measurement of this investigation includes air contents, slump loss, compressive strength, flexural strength, flexural toughness and neutron shielding rate. Four different fiber volume fractions were selected ranging from 0.25% to 1.0% by volume for the amorphous boron steel fibers. The testing results showed that the flexural toughness and the neutron shielding rate were increase with the increase of volume fraction for amorphous boron steel fiber. Based on the result, it is concluded that the concrete with the amorphous boron steel fiber can be effectively applied to shield the neutron and to improve mechanical properties.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• Journal of the Korean Society of Radiology
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• v.13 no.4
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• pp.565-572
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• 2019

In this study, shielding analysis of material and thickness of 3D printer filaments was performed for the manufacture of custom shielding by radiation workers during outdoor radiographic test. The shielding was attached to the ICRU Slab Phantom after selecting the voxel source $^{192}Ir$ and $^{75}Se$ through simulation using MCNPX, and the distance between the source and the slab Phantom was set at 100 cm. The 12 shielding materials were divided into 5 mm units up to 200 mm from the absence of shielding materials to evaluate the energy absorbed per unit mass of each shielding material. The results showed that the shielding effect was high in the order of ABS + Tungsten, ABS + Bismuth, PLA + Copper, PLA + Iron from all sources of radiographic test. However, compared to lead, the shielding effect was somewhat lower. Based on this study in the future, further study of the atomic number and the high density filament material is necessary.