• 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.

• 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 Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.41-50
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• 2006

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by $\sim20$ tons.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.153-162
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• 2005

In this Paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by ${\~}$20 tons.

• Progress in Medical Physics
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• v.14 no.2
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• pp.90-98
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• 2003

Purpose : A new virtual simulation technique for craniospinal irradiation (CSI) that uses a CT-simulator was developed to improve the accuracy of field and shielding placement as well as patient positioning. Materials and Methods : A CT simulator (CT-SIM) and a 3-D conformal radiation treatment planning system (3D-CRT) were used to develop CSI. The head and neck were immobilized with a thermoplastic mask while the rest of the body was immobilized with a Vac-Loc. A volumetric image was then obtained with the CT simulator. In order to improve the reproducibility of the setup, datum lines and points were marked on the head and body. Virtual fluoroscopy was performed with the removal of visual obstacles, such as the treatment table or immobilization devices. After virtual simulation, the treatment isocenters of each field were marked on the body and on the immobilization devices at the conventional simulation room. Each treatment fields was confirmed by comparing the fluoroscopy images with the digitally reconstructed radiography (DRR) and digitally composited radiography (DCR) images from virtual simulation. Port verification films from the first treatment were also compared with the DRR/DCR images for geometric verification. Results : We successfully performed virtual simulations on 11 CSI patients by CT-SIM. It took less than 20 minutes to affix the immobilization devices and to obtain the volumetric images of the entire body. In the absence of the patient, virtual simulation of all fields took 20 min. The DRRs were in agreement with simulation films to within 5 mm. This not only reducee inconveniences to the patients, but also eliminated position-shift variables attendant during the long conventional simulation process. In addition, by obtaining CT volumetric image, critical organs, such as the eyes and the spinal cord, were better defined, and the accuracy of the port designs and shielding was improved. Differences between the DRRs and the portal films were less than 3 m in the vertebral contour. Conclusion : Our analysis showed that CT simulation of craniospinal fields was accurate. In addition, CT simulation reduced the duration of the patient's immobility. During the planning process. This technique can improve accuracy in field placement and shielding by using three-dimensional CT-aided localization of critical and target structures. Overall, it has improved staff efficiency and resource utilization by standard protocol for craniospinal irradiation.

• Korean Journal of Digital Imaging in Medicine
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• v.15 no.2
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• pp.13-18
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• 2013

Purpose : Fluoroscopy equipment, depending on the type of changes that occur in the patient's position ESD and study the patient's scatter ray of ESD Practitioners considered a comparative analysis was to evaluate the correct dose. Materials and Methods : HITACHI four overtube type TU-8000 Flat Detector and Under tube C-Arm Philips' Multi Diagnost Eleva with Flat Detector type were measured by. Each devices is a measure of the patient's esd randophantom position in tabel unfors Xi multi funtion then fixed to the abdomen fluoroscopy and 10 seconds, spot was measured three times, practitioners of the incident surface dose by considering the patient's scatter ray of the table for each device in the average human stomach 21cm thickness acrylic phantom ($25cm{\times}25cm$) Place the practitioner position after position randophantom unfors Xi multi funtion in the thyroid and stomach 1 minute by a fixed one-time fluoroscopy and measured. Results : 10 seconds and the patient perspective of the c-arm ESD 1.2 times smaller on the AP and oblique measurements were measured in the 6-13 times smaller. spot positions to changes in the measured three times on the AP of the abdomen, ESD is 18 times smaller c-arm measurements and the oblique measurement was 19-30 times smaller. And 1 minute at practitioners fluoroscopy esd in the thyroid 2.12 times the c-arm, chest 1.75 times less the dose was measured. On the AP, depending on the device, but the lack of dose difference oblique positions of the two devices depending on changes in the area due to changes in both the AP than on the dose increased, the difference in dose between the two devices, the maximum difference was approximately 27 times. Conclusion : Fluoroscopic equipment at the time of inspection in accordance with changes in dose according to the patient and the patient's positions changes, because the area of the scatter ray considering the change of dose measurements be made, and study of the equipment according to the characteristics of the efficiency and the exposure of the patient and practitioner is considered smooth study equipment manufacturers that can be done is to build the system and think that is also important. Various fluoroscopy when you check future changes in many factors of change in dose for the equipment in the laboratory system by considering the scatter ray radiation shielding for the management to take advantage of reckless undertube have been utilized as more exposure Reduction activities can help is considered as the direction.

• Journal of the Korean Society of Radiology
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• v.14 no.3
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• pp.193-201
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• 2020

The purpose of this study is to evaluate the photon characteristics according to the material and thickness of the electrons incidented through a linear accelerator. The computer simulation design is a linear accelerator target consisting of a 2 mm thick tungsten single material and a 1.8 mm and 2.3 mm thick tungsten and copper composite material. In the research method, First, the behavior of primary particles in the target was evaluated by electron fluence and electron energy deposition. Second, photons occurring within the target were evaluated by photon fluence. Finally, the photon angle-energy distribution at a distance of 1 m from the target was evaluated by photon fluence. As a result, first, electrons, which are primary particles, were not released out of the target for electron fluence and energy deposition in the target of a single material and a composite material. Then, electrons were linearly attenuated negatively according to the target thickness. Second, it was found that the composite material target had a higher photon generation than the single material target. This confirmed that the material composition and thickness influences photon production. Finally, photon fluence according to the angular distribution required for shielding analysis was calculated. These results confirmed that the photon generation rate differed depending on the material and thickness of the linear accelerator target. Therefore, this study is necessary for designing and operating a linear accelerator use facility for container security screening that is being introduced in the country. In addition, it is thought that it can be used as basic data for radiation protection.

• Journal of the Korea Concrete Institute
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• v.26 no.3
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• pp.315-321
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• 2014

Concrete as a construction material is widely used in nuclear vessel and plant for excellent radiation shielding. However the isolation characteristics in concrete may affect adversely in the case of fire and melt-down in nuclear vessel since temperature cooling down is very difficult from outside. This study is for development of high thermal conductive concrete, and its mechanical and thermal properties are evaluated. Magnetite aggregates with volume ratio of 42.3% (maximum) and steel powder of 1.5% are replaced with normal aggregates and thermal properties are evaluated. Thermal conductivity little increases by 30% addition of magnetite but rapidly increases afterwards. Finally thermal conductivity is magnified to 2.5 times in the case of 42.3% addition of magnetite. Steel powder has a positive effect on high thermal conduction to 106~113%. Several models for thermal conduction like ACI, DEMM, and MEM are compared with test results and they are verified to reasonably predict the thermal conductivity with increasing addition of magnetite aggregates and steel powder.

• 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.