• Radiation Oncology Journal
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• v.11 no.1
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• pp.35-42
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• 1993

The dose response of the number of micronuclei in cytokinesis-blocked (CB) lymphocytes after in vitro irradiation with $\gamma$-rays and neutrons in the 5 dose ranges was studied for a heterogeneous population of 4 donors. One thousand binucleated cells were systematically scored for micronuclei. Measurements performed after irradiation showed a dose-dependent increase in micronuclei (MN) frequency in each of the donors studied. The dose-response curves were analyzed by a linear-quadratic model, frequencies per 1000 CB cells were ($0.31{\pm}0.049$)D+($0.0022{\pm}0.0002)D^2+(13.19{\pm}1.854) (r^2=1.000,\;X^2=0.7074,\;p=0.95$) following $\gamma$ irradiation, and ($0.99{\pm}0.528$)\;D+(0.0093{\pm}0.0047)\;D^2+(13.31{\pm}7.309)\;(r^2=0.996,\;X^2=7.6834,\;p=0.11) following neutrons irradiation (D is irradiation dose in cGy). The relative biological effectiveness (RBE) of neutrons compared with $\gamma$-rays was estimated by best fitting linear-quadratic model. In the micronuclei frequency between 0.05 and 0.8 per cell, the RBE of neutrons was $2.37{\pm}0.17$. Since the MN assay is simple and rapid, it may be a good tool for evaluating the $\gamma$-ray and neutron response.

• Nuclear Engineering and Technology
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• v.7 no.3
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• pp.191-198
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• 1975

A study was made on the neutron dosimetry in a mixed gamma-neutron field with LiF thermoluminescent dosimeter. In order to estimate the neutron dose in a mixed field, $^{6}$ LiF and $^{7}$ LiF dosimeters were used for fast and thermal neutron doses. The over-all conversion factors for the effects of dosimeter positions were derived for personnel monitoring and the glow curves of the LiF dosimeters for neutron and gamma-ray doses were also analyzed.

• Nuclear Engineering and Technology
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• v.4 no.1
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• pp.39-45
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• 1972

In neutron shielding, the scattering effect is equally important as the attenuations in shielding materials. In the present study, the scattered dose equivalent was measured using a Rem counter for water, paraffin, borated paraffin, ordinary and heavy concrete, lead, iron, and tissue equivalent material in three different angles: 45$^{\circ}$, 90$^{\circ}$, and 135$^{\circ}$, respectively. The measurements were performed for the neutron, having the energies of 0.5, 1, 2, 5, and 18 MeV, which are produced from the Van do Graaff accelerator. The scattered dose equivalent ratios were increased with increasing the thickness of scattering materials and saturated at a certain thickness although they were different from one to other materials under study. The ratios were large for lead and iron while they were small for the hydrogen containing materials such as water and paraffin etc.

• 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.22 no.4
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• pp.289-298
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• 1997

Time-consuming experiments have been required in the development of neutron moisture gauge to induce a relation between the water content in soil and the neutron counts. Applying a monte carlo computer code to simulate the experiments of neutron moisture gauging may contribute to reduce time and efforts for experiments and produce a calibration equation which is more applicable to soil in general. In this study MCNP4A, a monte carlo computer code, was employed to simulate soil experiments and the simulated results were compared with experimental ones. The comparative study showed that MCNP4A is applicable to simulate the experiments and calibration equation can be obtained through simulations. Effects of dry density changes were also studied.

• Journal of the Korean Society of Radiology
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• v.9 no.5
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• pp.295-300
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• 2015

There are various type of particle accelerators such as Kyoungju 100-MeV proton beam accelerator in Korea. And Korea plans to build large particle accelerator such as heavy ion accelerator and 4th generation light source facility. The accelerated high energy particles of these facility produce 2nd neutron after nuclear reaction with target materials. And then these 2nd neutron activate structural materials and surrounding environment. Accordingly, it is important to consider the activation and shielding calculation on design of facility for safety operation. In this study, we tried to calculate and compare the neutron flux from the interaction $^{la}150$ beam with target material(Cu) according to thickness of iron and concrete shielding material by MCNPX 2.7 with nuclear library JENDL/HE 07and la150. To verify the properties of nuclear library, we compared computational results with experimental value. These results can be used for dose evaluation technology in planning of the shielding of large particle accelerator.

• Journal of Radiation Protection and Research
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• v.13 no.1
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• pp.42-54
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• 1988

Overall reviews of papers presented at the seventh IRPA International Congress (Aprill 10-17, 1988) held in Sydney, Australia have been done in order to grasp the trends of radiation research. In this report, the changing and increasing matters in the field of ionizing radiation safety as well as non-ionizing radiation application are introduced to the KARP. In addition, a research paper, 'Emission Rate Measurement of a Cf-252 Neutron Source by Manganous Sulfate Bath Method', presented at the IRPA 7 Poster Session is followed.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.15 no.6
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• pp.3734-3740
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• 2014

The aim of this study was provide basic information and establish the criteria in radiation therapy planning by measuring the absorbed neutron dose of normal tissues and lesions according to the number of portals. From September 2013 to January 2014, 20 patients who were diagnosed with prostate cancer and were previously treated with radiation therapy were replanned retrospectively to measure the absorbed neutron dose distribution according to the number of portals. The absorbed neutron dose was measured in each of the 5, 7 and 9 portals using a 15 MV energy, which meant a therapeutic dose of 220 cGy. The optical stimulation luminescence dosimeter was separated by 20cm and 60cm away from the center of the field of view. As a result, the average radiation dose in the abdomen appeared to have a positive relationship with the number of portals, which was statistically significant (p<.05). The average radiation dose was $4.34{\pm}1.08$. The average radiation dose in the thyroid was $2.71{\pm}.37$. Although it showed a positive relationship with the number of portals, it did not have statistical significance. The number of portals and the neutron dose depending on the position showed a significant positive relationship, particularly in the abdomen. As a result of linear regression analysis, as the number of the portal increased in steps, the average volume of the neutrons increased significantly (0.416 times). In conclusion, efficient selection of the number of portals is needed considering the difference in the absorbed neutron dose in the normal tissues depending on the number of the portals.

• Journal of Radiation Protection and Research
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• v.21 no.4
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• pp.297-311
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• 1996

ANSI decided PMMA slab phantom as a calibration phantom and introduced a conversion coefficient calculation method for it. For photon, the conversion coefficient can be obtained by using backscatter factor and conversion coefficient of the ICRU tissue cube and backscatter factor of the PMMA slab. For neutron, however, the ANSI has not introduced any conversion coefficient calculation method for the PMMA slab. In this work, the ANSI method for the photon conversion coefficient calculation was applied to the neutron conversion coefficient calculation of the PMMA slab. Quality weighted tissue kerma of neutron was applied to calculate the backscatter factors on the ICRU cube and the PMMA slab. The dose conversion coefficient of the ICRU cube was also calculated by using MCNP code. Then, the dose conversion coefficient of the PMMA slab was calculated from two backscatter factors and the dose conversion coefficient of the ICRU cube. The discrepancies of the dose conversion coefficients of the PMMA slab and the ICRU cube were less than 10% except 1eV(20%), 1keV(17%), and 4 MeV(16%).

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.305-305
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• 2004

핵물질을 취급하는 시설에서는 핵물질 안전조치 목적의 달성, 즉 핵물질의 군사적 전용 및 도난을 방지하기 위한 하나의 수단으로서 핵물질의 취급 및 이동을 감시하기 위한 감시시스템이 요구된다. 이 연구에서는 이러한 요구에 부응하기 위해 시설 내에서 핵물질이 이동 가능한 모든 경로에 중성자 모니터와 카메라 같은 감시 장비를 설치하고, 이들로부터 실시간으로 방사선 신호와 영상 데이터를 취득ㆍ분석하여 핵물질의 거동을 진단할 수 있는 핵물질 감시시스템을 개발하였다.(중략)