• Journal of Radiation Protection and Research
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• v.36 no.4
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• pp.200-208
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• 2011

The neutron fluence measurement and evaluation technology is very important for material irradiation test. The most essential technology in this study is the neutron irradiation evaluation method using a fluence monitor. The fluence monitors were fabricated with metal wires of the purity ${\geq}$ 99.9%, whose dimensions were 0.1mm diameter, about 3 mm length, and around 150-200 ${\mu}g$ mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. After irradiation tests, radiation activities were measured with the high purity germanium (HPGe) detector. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position.

• Journal of Radiation Protection and Research
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• v.29 no.2
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• pp.73-90
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• 2004

The Reg. Guide 1.109 model was reviewed against its applicability to calculating radionuclide concentrations in agricultural products for operating nuclear facilities and an improved method was proposed. The model was so modified that the radionuclides deposited since the start of operation could be considered in assessing the root uptake. Translocation factors were introduced in the equation for calculating the concentrations in edible parts due to direct plant deposition. Values specific to Korea were set up for the input parameters of the modified model. The concentrations of $^{54}Mn,\;^{60}Co,\;^{90}Sr\;and\;^{137}Cs$ in rice seeds, Chinese cabbage and radish root were calculated for various hypothetical deposition histories using the Reg. Guide 1.109 model and the modified model with parameter values in the guide and those specific to Korea put in alternately. Through comparisons among the results, it could be expected that the use of the modified model with the input of parameter values specific to Korea would result In a more resonable and realistic assessment.

• Journal of Radiation Protection and Research
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• v.6 no.1
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• pp.8-24
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• 1981

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute(ANSI) N666. These data are used to calculated the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from $2.5{\times}10^{-8}$ to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoetiergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be a useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions.

• Journal of Conservation Science
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• v.16 s.16
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• pp.5-14
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• 2004

Luminescence dating is based upon the premise that several commonly occurring minerals (e.g. quartz and feldspar) can be used as natural dosimeters, recording the amount of radiation to which they have been exposed. We report results of optical dating on quartz samples separated from archaeological remains (burnt soils and potteries), which were excavated at Myungam-Ri, Asan, and Chungnam province and at Sayang-Ri, Chinchun, Chungbuk province, considered as the cultural site of the early Bronze Age. The resultant dates were in good agreement with the ages derived by archaeological assessment. It is concluded that the optical dating introduced in here should contribute significantly to future archaeological dating work.

• Radiation Oncology Journal
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• v.2 no.1
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• pp.129-137
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• 1984

In brachytherapy, it is important to determine the positions of the radiation sources which are inserted into a patient and to estimate the dose resulting from the treatment. Calculation of the dose distribution throughout an implant is so laborious that it is rarely done by manual methods except for model cases. It is possible to calculate isodose distributions and tumor doses for individual patients by the use of a microcomputer. In this program, the dose rate and dose distributions are calculated by numerical integration of point source and the localization of radiation sources are obtained from two radiographs at right angles taken by a simulator developed for the treatment planning. By using microcomputer for brachytherapy, we obtained the result as following 1. Dose calculation and irradiation time for tumor could be calculated under one or five seconds after input data. 2. It was same value under$\pm2\%$ error between dose calculation by computer program and measurement dose. 3. It took about five minutes to reconstruct completely dose distribution for intracavitary irradiation. 4. Calculating by computer made remarkly reduction of dose errors compared with Quimby's calculation in interstitial radiation implantation. 5. It could calculate the biological isoffect dose for high and low dose rate activities.

• Proceedings of the Korea Information Processing Society Conference
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• 2005.11a
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• pp.1403-1406
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• 2005

원자력 시설 및 연구용 원자로 해체 시 해체 일정과 해체 폐기물량 그리고 해체 비용을 분석하기 위한 평가식이 수립되었다. 연구로 2 호기 Thermal column 자료를 참고하여 평가식을 실험하였다. 해체 공정 모사 결과를 애니메이션으로 보여주는 가시화 모듈과 해체 일정과 해체 폐기물량, 작업자 피폭선량 및 해체 비용 등을 평가식으로 계산 한 후 그 결과를 그래픽으로 보여주는 시뮬레이션 모듈로 구성된 해체 디지털 목업 시스템의 그래픽 사용자 인터페이스가 설계되었다. 해체 단위 작업별 평가식은 원자력과 관련한 시설 해체 시 해체 일정 및 해체 비용 분석 및 예측에 중요한 기초자료로 사용 될 것이다. 또한 그래픽 사용자 인터페이스는 방사능의 오염으로 인해 작업자가 접근하기 힘든 환경에서의 해체 활동을 사전에 경험함으로써 피폭으로부터 작업자의 안전성을 향상시킬 수 있는 유용한 도구로 활용될 수 있다.

• Journal of radiological science and technology
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• v.23 no.2
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• pp.75-79
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• 2000

This study was performed for the clinical applications applying the Monte Carlo methods. In this study we calculated the absorbed dose distributions for the 6 MeV electron beam in water phantom and compared the results with measured values. The energy data of electron beam used in Monte Carlo calculation is the energy distribution for 6 MeV electron beam which is assumed as a Gaussian form. We calculated percent depth doses and beam profiles for three field sizes of $10{\times}10,\;15{\times}15$, and $20{\times}20\;cm^2$ in water phantom using Monte Carlo methods and measured those data using a semiconductor detector and other devices. We found that the calculated percent depth doses and beam profiles agree with the measured values approximately. However, the calculated beam profiles at the edge of the fields were estimated to be lower than the measured values. The reason for that result is that we did not consider the angular distributions of the electrons in phantom surface and contamination of X-rays in our calculations. In conclusion, in order to apply the Monte Carlo methods to the clinical calculations we are to study the source models for electron beam of the linear accelerator beforehand.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.353-363
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• 2006

In order to evaluate the potential radiological risks for the reuse of the site after decommissioning of nuclear facilities, a mathematical model was developed and materialized into the Microsoft $Excel{\circledR}$ spreadsheets frame. A set of input parameter values was proposed, which is useful in the preliminary risk screening step before the detailed evaluation with the site-specific data. It appeared that the screening levels calculated by the present model was agreed with the derived concentration guideline limits resulted from RESRAD Ver.6.2 and the German dose criteria for releasing a nuclear site from regulatory control.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.596-600
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• 2003

The Advanced spent fuel Conditioning Process(ACP) is under development for the effective management of spent fuel which had been generated in nuclear plants. The ACP needs a hot cell where most operations will be peformed. To give priority to the environment safety, radiation doses evaluation for the radioactive nuclides were preliminarily peformed in both normal operation and accident case. The evaluation result shows a safe margin for regulation limits and SAR limit of IMEF where this facility will be constructed.

• Progress in Medical Physics
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• v.14 no.2
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• pp.69-73
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• 2003

In this study, we measured shielding block correction factors for irregular fields and compared them with published data for the square blocked field. We devised a methods to measure the factors at an arbitrary depth in phantom. The measurements were performed for 12 shielding blocks used in radiation therapy. The measured correction factors for irregular blocked fields were consistent within $\pm$0.5% with those of the square blocked fields. Our results show that the shielding block correction factors for the typical square blocked fields can be used in clinical dose calculations for irregular blocked fields. However, for small fields, we suggest that verification be done by measurement.