• Journal of the Korean Society of Radiology
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• v.9 no.4
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• pp.235-238
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• 2015

The aims of this study were to evaluate the difference of renal uptake rate in $^{99m}Tc-DMSA$ scan on pediatrics by including the bladder. Phantom and Clinical studies were performed. In the phantom study, we put $^{99m}TcO_4{^-}$ (300uCi, 11 MBq) in 3cups filled with distilled water at the rate 1:1:0, 1:1:0.5, 1:1:1, 1:1:2 and were placed Lt kidney, Rt kidney and bladder position on the table. To acquire the image, we used Symbia-E gamma camera from Siemens with preset count method(400,000 counts). In quantitative analysis, the counts of drawing ROIs on the phantom were analyzed. In clinical studies, we analyzed the 20 pediatrics who were examined by $^{99m}Tc-DMSA$ scan. At first, the images were acquired with both kidney and bladder. Secondly we acquired images after shielding the bladder. And the data were compared using a pared t-test by SPSS(ver.22.0). As a result of renal phantom's experiment, we compared with average of uptake rate(%), 1:1:0 was Lt 43.32%, Rt 45.97%, 1:1:0.5 was Lt 35.79%, Rt 36.89%, 1:1:1 was Lt 29.68%, Rt 31.45% and 1:1:2 was Lt 22.89%, Rt 24.32%. There was no correlation between the zoom and uptake rate. The results of patient were that excluded bladder was $29.83{\pm}8.81%$(Lt), $24.29{\pm}6.66%$(Rt) and included bladder was $26.65{\pm}8.03%$(Lt, $21.78{\pm}6.24%$(Rt). This is deemed statistically significant (p<0.05). Renal uptake rate was undervalued because the counts of bladder were included in the total counts.

• Journal of the Korean Society for Nondestructive Testing
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• v.27 no.1
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• pp.31-37
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• 2007

A $^{192}Ir$ small-focal source has been developed by using the HANARO reactor and the radioisotope production facility at the Korea Atomic Energy Research Institute (KAERI). The small-focal source with the dimension of 0.5 mm in diameter and 0.5 mm in length was fabricated as an aluminum-encapsulated form by a specially designed pressing equipment. For the estimation of the radioactivity, neutron self-shielding and ${\gamma}-ray$ self-absorption effects on the measured activity was considered. From this estimation, it is realized that $^{192}Ir$ small-focal sources over 3 Ci activities can be produced from the HANARO. Field performance tests were performed by using a conventional source and the developed source to take images of a computer CPU and a piece of a carbon steel. The small-focal source showed better penetration sensitivity and geometrical sharpness than the conventional source does. It is concluded from the tests that the focal dimension of this source is small enough to maximize geometrical sharpness in the image taking for the close proximity shots, pipeline crawler applications and contact radiography.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.503-510
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• 2020

Concrete is one of the most widely used materials as the shielding structures of a nuclear facilities. It is also the most generated radioactive waste in quantity while dismantling facilities. Since the concrete captures neutrons and generates various radionuclides, radiation measurement and analysis of the sample was fulfilled prior to dismantle facilities. An HPGe detector is used in general for the radiation measurement, and effective correction factors such as geometrical correction factor, self-absorption correction, and absolute detector efficiency have to be applied to the measured data to decide exact radioactivity of the sample. Correction factors are obtained by measuring data using a standard source with the same geometry and chemical states as the sample under the same measurement conditions. However, it is very difficult to prepare standard concrete sources because concrete is limited in pretreatment due to various constituent materials and high density. In addition, the concrete sample obtained by core drill is a volumetric source, which requires geometric correction for sample diameter and self absorption correction for sample density. Therefore in recent years, many researchers are working on the calculation of effective correction factors using Monte carlo simulation instead of measuring them using a standard source. In this study we calculated, using Geant4, one of the Monte carlo codes, the correction factors for the various diameter and density of the concrete core sample at the gamma ray energy emitted from the nuclides ¹⁵²Eu and ⁶⁰Co, which are the most generated in radioactive concrete.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.90-96
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• 1984

Estimated dose rates in spent fuel pool storage with the extended fuel cycle core management were reviewed and compared with design limit after calculation with the aid of DLC-23/CASK(22 n, 18 g) nuclear data and ANISN code. Radioactivity and gamma spectrum within spent fuel assemblies were calculated with ORIGEN code by extended fuel cycle model. In the calculation of dose rate, the fuel pool geometry was assumed to be infinite slab. Also, composition materials and radiation source within assemblies which are being stored in pool storage were assumed to be uniformly distributed throughout all the assemblies. As a result of culculation of dose rate from stored assemblies and waterborne radionuclides in pool water, the calculated dose rates appear to be lower than design basis limit under normal condition as well as abnormal condition.

• The Korean Journal of Nuclear Medicine Technology
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• v.15 no.1
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• pp.75-80
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• 2011

Purpose: Europe and U.S use multi-bed therapy rooms. Hereupon, this study aims to examine the safety when current one-bed therapy rooms in Seoul National University Hospital is changed into two-bed ones. Materials and Methods: This study evaluated external exposure by gamma radiation emitted from other patients and internal and external exposure caused by pollutions from other patients in case that Seoul National University Hospital installs a shielding wall between beds in existing therapy rooms. Results: When internal and external exposure was evaluated to evaluate safety of two-bed hospital rooms, 'isolation amount of patients' 5mSv exposure or below is received according to the Atomic Energy Act. Conclusion: With the increasing number of patients with thyoid cancer, patients using therapy rooms are on the rise. Therefore, improving one-person therapy rooms to two-person ones in line with international trend would increase cost reduction and management efficiency, and patients' alienation and isolation can be reduced to increase healing effects.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.243-260
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• 2018

A literature review on the effects of high temperature and radiation on radiation shielding concrete in Spent Fuel Dry Storage is presented in this study with a focus on concrete degradation. The general threshold is $95^{\circ}C$ for preventing long-term degradation from high temperature, and it is suggested that the temperature gradient should be less than $60^{\circ}C$ to avoid crack generation in concrete structures. The amount of damage depends on the characteristics of the concrete mixture, and increases with the temperature and exposure time. The tensile strength of concrete is more susceptible than the compressive strength to degradation due to high temperature. Nuclear heating from radiation can be neglected under an incident energy flux density of $10^{10}MeV{\cdot}cm^{-2}{\cdot}s^{-1}$. Neutron radiation of >$10^{19}n{\cdot}cm^{-2}$ or an integrated dose of gamma radiation exceeding $10^{10}$ rads can cause a reduction in the compressive and tensile strengths and the elastic moduli. When concrete is highly irradiated, changes in the mechanical properties are primarily caused by variation in water content resulting from high temperature, volume expansion, and crack generation. It is necessary to fully utilize previous research for effective technology development and licensing of a Korean dry storage system. This study can serve as important baseline data for developing domestic technology with regard to concrete casks of an SF (Spent Fuel) dry storage system.

• The Korean Journal of Nuclear Medicine Technology
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• v.20 no.1
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• pp.32-36
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• 2016

Purpose In nuclear medicine examination, $^{131}I$ is widely used in nuclear medicine examination such as diagnosis, treatment, and others of thyroid cancer and other diseases. $^{131}I$ conducts examination and treatment through emission of ${\gamma}$ ray and ${\beta}^-$ ray. Since $^{131}I$ (364 keV) contains more energy compared to $^{99m}Tc$ (140 keV) although it displays high integrated rate and enables quick discharge through kidney, the objective of this study lies in comparing the difference in exposure dose of $^{131}I$ before and after wearing apron when handling $^{131}I$ with focus on 3 elements of external exposure protection that are distance, time, and shield in order to reduce the exposure to technicians in comparison with $^{99m}Tc$ during the handling and administration process. When wearing apron (in general, Pb 0.5 mm), $^{99m}Tc$ presents shield of over 90% but shielding effect of $^{131}I$ is relatively low as it is of high energy and there may be even more exposure due to influence of scattered ray (secondary) and bremsstrahlung in case of high dose. However, there is no special report or guideline for low dose (74 MBq) high energy thus quantitative analysis on exposure dose of technicians will be conducted based on apron wearing during the handling of $^{131}I$. Materials and Methods With patients who visited Department of Nuclear Medicine of our hospital for low dose $^{131}I$ administration for thyroid cancer and diagnosis for 7 months from Jun 2014 to Dec 2014 as its subject, total 6 pieces of TLD was attached to interior and exterior of apron placed on thyroid, chest, and testicle from preparation to administration. Then, radiation exposure dose from $^{131}I$ examination to administration was measured. Total procedure time was set as within 5 min per person including 3 min of explanation, 1 min of distribution, and 1 min of administration. In regards to TLD location selection, chest at which exposure dose is generally measured and thyroid and testicle with high sensitivity were selected. For preparation, 74 MBq of $^{131}I$ shall be distributed with the use of $2m{\ell}$ syringe and then it shall be distributed after making it into dose of $2m{\ell}$ though dilution with normal saline. When distributing $^{131}I$ and administering it to the patient, $100m{\ell}$ of water shall be put into a cup, distributed $^{131}I$ shall be diluted, and then oral administration to patients shall be conducted with the distance of 1m from the patient. The process of withdrawing $2m{\ell}$ syringe and cup used for oral administration was conducted while wearing apron and TLD. Apron and TLD were stored at storage room without influence of radiation exposure and the exposure dose was measured with request to Seoul Radiology Services. Results With the result of monthly accumulated exposure dose of TLD worn inside and outside of apron placed on thyroid, chest, and testicle during low dose $^{131}I$ examination during the research period divided by number of people, statistics processing was conducted with Wilcoxon Signed Rank Test using SPSS Version. 12.0K. As a result, it was revealed that there was no significant difference since all of thyroid (p = 0.345), chest (p = 0.686), and testicle (p = 0.715) were presented to be p > 0.05. Also, when converting the change in total exposure dose during research period into percentage, it was revealed to be -23.5%, -8.3%, and 19.0% for thyroid, chest, and testicle respectively. Conclusion As a result of conducting Wilcoxon Signed Rank Test, it was revealed that there is no statistically significant difference (p > 0.05). Also, in case of calculating shielding rate with accumulate exposure dose during 7 months, it was revealed that there is irregular change in exposure dose for inside and outside of apron. Although the degree of change seems to be high when it is expressed in percentage, it cannot be considered a big change since the unit of accumulated exposure dose is in decimal points. Therefore, regardless of wearing apron during high energy low dose $^{131}I$ administration, placing certain distance and terminating the administration as soon as possible would be of great assistance in reducing the exposure dose. Although this study restricted $^{131}I$ administration time to be within 5 min per person and distance for oral administration to be 1m, there was a shortcoming to acquire accurate result as there was insufficient number of N for statistics and it could be processed only through non-parametric method. Also, exposure dose per person during lose dose $^{131}I$ administration was measured with accumulated exposure dose using TLD rather than through direct-reading exposure dose thus more accurate result could be acquired when measurement is conducted using electronic dosimeter and pocket dosimeter.

• The Journal of Korean Society for Radiation Therapy
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• v.30 no.1_2
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• pp.177-183
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• 2018

Purpose : Volumetric Modulated Arc Therapy(VMAT) has the advantage of uniformly and precisely irradiating the tumor to the shape of the tumor while reducing the risk of radiation damage to normal tissues. such as brain cancer, head and neck cancer and prostate cancer, It is being used for treatment. The purpose of this study is to evaluate the usefulness of the Jaw-Tracking technique(JTT) in VMAT for brain and head and neck cancer. Materials and Methods : We selected eight patients with brain and head and neck cancer(4 Brain, 4 head and neck) who were treated with the VMAT treatment technique. Contouring information of the patient's tumor and normal organ was fused to the Rando phantom using the deformable registration of Velocity(Varian, USA). A treatment plan was developed using the Varian Eclipse(ver 15.5, Varian, USA) with the same patient actual beam parameters except for the use of jaw-tracking. As the evaluation index, the maximum dose and mean dose of target and OAR were compared and a portal dosimetry was performed for the treatment plan verification. Results : When using JTT, the relative dose of OAR decreased by 5.24 % and the maximum dose by 7.05 %, respectively, compared with the Static-Jaw technique(SJT). In the various OARs, the mean dose and maximum dose reduction ranges ranged from 0.01 to 3.16 Gy and from 0.12 to 6.27 Gy, respectively. In the case of the target, the maximum dose of GTV, CTV, PTV decreased by 0.17 %, 0.43 %, and 0.37 % in JTT, and the mean dose decreased by 0.24 %, 0.47 % and 0.47 %, respectively. Gamma analysis The JTT and SJT passing rates were $98{\pm}1.73%$ and $97{\pm}1.83%$ on the basis of 3 % / 3 mm, respectively. Comparing the doses of all OARs applied to the experiment, it was found that the use of JTT resulted in a significant decrease in dose due to additional jaw shielding besides MLC than SJT. Conclusion : In radiation therapy using VMAT treatment plan, we can apply JTT in the case of adjacent tumor and normal organs such as brain cancer and head and neck cancer, and in radiotherapy required large field and high energy caused increase leakage dose through MLC. It is considered that the target dose of PTV can be increased by lowering the dose of normal tissue surrounding the tumor.