• Progress in Medical Physics
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• v.18 no.2
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• pp.87-92
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• 2007

A thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the HANARO, 30 MW multi-purpose research reactor. Mixed beams with different physical characteristics and relative biological effectiveness would be emitted from the BNCT irradiation facility, so a quantitative analysis of each component of the mixed beams should be performed to determine the accurate delivered dose. Thus, various techniques were applied including the use of activation foils, TLDs and ionization chambers. All the dose measurements were perform ed with the water phantom filled with distilled water. The results of the measurement were compared with MCNP4B calculation. The thermal neutron fluxes were $1.02E9n/cm^2{\cdot}s\;and\;6.07E8n/cm^2{\cdot}s$ at 10 and 20 mm depth respectively, and the fast neutron dose rate was insignificant as 0.11 Gy/hr at 10 mm depth in water The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water Good agreement within 5%, has been obtained between the measured dose and the calculated dose using MCNP for neutron and gamma component and discrepancy with 14% for fast neutron flux Considering the difficulty of neutron detection, the current study support the reliability of these results and confirmed the suitability of the thermal neutron beam as a dosimetric data for BNCT clinical trials.

• Proceedings of the KIEE Conference
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• 2005.10b
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• pp.430-432
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• 2005

저에너지 중성자가 가톨리늄(Gd) 막에 입사되면 중성자 포획과정에서 전환전자가 생성된다. 이 전환전자에 의해 pMOSFET $SiO_2$ 산화층에서 발생된 전자-전공쌍이 발생되고, 이 가운데 정공은 산화층 내부에 쉽게 붙잡혀(Trap) 양전하 센터로 작용하게 된다. 이 축적된 전하는 pMOSFET의 문턱전압(Threshold voltage)을 변화시킨다. 본 연구에서는 이러한 간접측정 원리를 이용하여 열중성자를 실기간 탐지할 수 있는 반도체형 탐지소자를 개발하고 하나로(HANARO) 방사선장에서의 시험을 통해 성능을 검증하였다. 그리고 감도관련 변수의 최적화를 통하여 작업자가 사용 가능한 범위의 고감도 열중성자 선량계로 개선 제작하였다. 개발된 선량계는 소형으로 실시간 열중성자 측정이 가능하며 감마방사선으로부터 독립적으로 열중성자를 측정할 수 있는 장점도 지니고 있다.

• Radiation Oncology Journal
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• v.24 no.4
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• pp.280-284
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• 2006

$\underline{Purpose}$: To evaluate biological characteristics of neutron beam generated by MC50 cyclotron located in the Korea Institute of Radiological and Medical Sciences (KIRAMS). $\underline{Materials\;and\;Methods}$: The neutron beams generated with 15 mm Beryllium target hit by 35 MeV proton beam was used and dosimetry data was measured before in-vitro study. We irradiated 0, 1, 2, 3, 4 and 5 Gy of neutron beam to EMT-6 cell line and surviving fraction (SF) was measured. The SF curve was also examined at the same dose when applying lead shielding to avoid gamma ray component. In the X-ray experiment, SF curve was obtained after irradiation of 0, 2, 5, 10, and 15 Gy. $\underline{Results}$: The neutron beams have 84% of neutron and 16% of gamma component at the depth of 2 cm with the field size of $26{\times}26\;cm^2$, beam current $20\;{\mu}A$, and dose rate of 9.25 cGy/min. The SF curve from X-ray, when fitted to linear-quadratic (LQ) model, had 0.611 as ${\alpha}/{\beta}$ ratio (${\alpha}=0.0204,\;{\beta}=0.0334,\;R^2=0.999$, respectively). The SF curve from neutron beam had shoulders at low dose area and fitted well to LQ model with the value of $R^2$ exceeding 0.99 in all experiments. The mean value of alpha and beta were -0.315 (range, $-0.254{\sim}-0.360$) and 0.247 ($0.220{\sim}0.262$), respectively. The addition of lead shielding resulted in no straightening of SF curve and shoulders in low dose area still existed. The RBE of neutron beam was in range of $2.07{\sim}2.19$ with SF=0.1 and $2.21{\sim}2.35$ with SF=0.01, respectively. $\underline{Conclusion}$: The neutron beam from MC50 cyclotron has significant amount of gamma component and this may have contributed to form the shoulder of survival curve. The RBE of neutron beam generated by MC50 was about 2.2.

• Radiation Oncology Journal
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• v.19 no.1
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• pp.66-73
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• 2001

Purpose : For the research of Boron Neutron Capture Therapy (BNCT), fast neutrons generated from the MC-50 cyclotron with maximum energy of 34.4 MeV in Korea Cancer Center Hospital were moderated by 70 cm paraffin and then the dose characteristics were investigated. Using these results, we hope to establish the protocol about dose measurement of epi-thermal neutron, to make a basis of dose characteristic of epi-thermal neutron emitted from nuclear reactor, and to find feasibility about accelerator-based BNCT. Method and Materials : For measuring the absorbed dose and dose distribution of fast neutron beams, we used Unidos 10005 (PTW, Germany) electrometer and IC-17 (Far West, USA), IC-18, ElC-1 ion chambers manufactured by A-150 plastic and used IC-l7M ion chamber manufactured by magnesium for gamma dose. There chambers were flushed with tissue equivalent gas and argon gas and then the flow rate was S co per minute. Using Monte Carlo N-Particle (MCNP) code, transport program in mixed field with neutron, photon, electron, two dimensional dose and energy fluence distribution was calculated and there results were compared with measured results. Results : The absorbed dose of fast neutron beams was $6.47\times10^{-3}$ cGy per 1 MU at the 4 cm depth of the water phantom, which is assumed to be effective depth for BNCT. The magnitude of gamma contamination intermingled with fast neutron beams was $65.2{\pm}0.9\%$ at the same depth. In the dose distribution according to the depth of water, the neutron dose decreased linearly and the gamma dose decreased exponentially as the depth was deepened. The factor expressed energy level, $D_{20}/D_{10}$, of the total dose was 0.718. Conclusion : Through the direct measurement using the two ion chambers, which is made different wall materials, and computer calculation of isodose distribution using MCNP simulation method, we have found the dose characteristics of low fluence fast neutron beams. If the power supply and the target material, which generate high voltage and current, will be developed and gamma contamination was reduced by lead or bismuth, we think, it may be possible to accelerator-based BNCT.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1949-1958
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• 2023

The International Commission on Radiological Protection (ICRP) Publication 116 was released to provide a comprehensive dataset of the dose coefficients (DCs) for external exposures produced with the adult reference voxel phantoms of ICRP Publication 110. Although an advanced skeletal dosimetry method for photons and neutrons using fluence-to-dose response functions (DRFs) was introduced in ICRP Publication 116, the ICRP-116 skeletal DCs were calculated by using the simple method conventionally used (i.e., doses to red bone marrow and endosteum approximated by doses to spongiosa and/or medullary cavities). In the present study, the photon and neutron DRFs were used to produce skeletal DCs of the ICRP-110 reference phantoms, which were then compared with the ICRP-116 DCs. For photons, there were significant differences by up to ~2.8 times especially at energies <0.3 MeV. For neutrons, the differences were generally small over the entire energy region (mostly <20%). The general impact of the DRF-based skeletal DCs on the effective dose calculations was negligibly small, supporting the validity of the ICRP-116 effective DCs despite their skeletal DCs derived from the simple method. Meanwhile, we believe that the DRF-based skeletal DCs could be beneficial in better estimates of skeletal doses of individuals for risk assessments.

• Journal of Radiation Protection and Research
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• v.2 no.1
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• pp.17-23
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• 1977

The neutron dosimetrical parameters, i. e., the fission neutron spectrum-averaged cross-sections and the fluence-to-dose conversion factors have been calculated for some threshold detectors with the aid of a computer. The threshold detectors under investigation were the $^{115}In(n,\;n')^{115m}In,\;^{32}S(n,\;p)^{32}P$ and $^{27}Al(n,\;{\alpha})^{24}Na$ reactions. It is revealed that the average cross-sections($\bar{\sigma}$) for the $^{32}S(n,\;p)^{32}P$ reaction are independent of the spectral functions, namely, the Watt-Cranberg and Maxwellian forms. In the case of the $^{27}Al(n,\;{\alpha})^{24}Na$ reaction a variation of the $\bar{\sigma}$ values appears to be highly dependent on the fissioning types. It seems that both the average cross-section for the $^{115}In(n,\;n')^{11m}In$ reaction and the conversion factor are insensitive to the spectral deformation of fission neutrons. These phenomena make it applicable to use indium as a possible integral fast neutron dosimeter in nuclear criticality accidents provided that the virgin fission neutrons are completely free from the scattered neutrons.

• Radiation Oncology Journal
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• v.6 no.1
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• pp.85-91
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• 1988

For the physical characterization of neutron beam, dosimetric measurements had been performed to obtain physical data of KCCH cyclotron-produced neutrons for clinical use. The results are presented and compared with the data of other institutions from the literatures. The central axis percent depth dose, build-up curves and open and wedge isodose curve values are intermediate between that of a 4 and 6 MV X-rays. The build-up level of maximum dose was at 1.35cm and entrance dose was approximately $40\%$. Flatness of the beam was $9\%$ at Dmax and less $than{\pm}3\%$ at the depth of $80\%$ isodose line. Penumbra begond the $20\%$ line is wider than corresponding photon beam. The output factors ranged 0.894 for $6\times6cm$ field to 1.187 for $30\times30cm$ field. Gamma contamination of neutron beam was $4.9\%$ at 2 cm depth in $10\times10cm$ field.

• Journal of Radiation Protection and Research
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• v.21 no.2
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• pp.131-138
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• 1996

This paper describes the methods and results of the personnel dosimetry performance tests which were been implemented for the first time in Korea in 1995. Seven categories, except the neutron category prescribed in the ANSI N13.11-1993, were adopted in the test. Fifteen types of dosimeters were participated by fourteen dosimeter processing institutes. A total of 129 dosimeters were selected to test-each type - 15 dosimeters for each of the seven categories and 24 for the controls. A total of 144 radiation categories were employed in the test and a total of 2560 (including 400 controls)dosimeters were submitted-7 categories for each type of the fifteen types dosimeters and 39 categories for the retest. The performance index in each category. sum of the absolute value of the bias and the standard deviation value of the performance quotient. was estimated by the use of delivered and processed dose equivalents according to the standard procedure. The performance in a given category was assessed as acceptable, for the deep and shallow dose equivalents (or the absorbed dose), if the performance index was less than 0.5. The test results showed 54% of the processors passed in the first test, 33% in the retest and 13% in the second retest.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.1220-1224
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• 2003

In this paper, we measured the radiation dose of a fuel handling machine of the CANDU type Wolsong nuclear reactor directly during operation, in spite of the high radiation level. In this paper we will describe the sensor development, measurement techniques, and results of our study. For this study, we used specially developed semiconductor sensors and matching dosimetry techniques for the mixed radiation field. MOSFET dosimeters with a thin oxide, that are tuned to a high dose, were used to measure the ionizing radiation dose. Silicon diode dosimeters with an optimum area to thickness ratio were used for the radiation damage measurements. The sensors are able to distinguish neutrons from gamma/X-rays. To measure the radiation dose, electronic sensor modules were installed on two locations of the fuel handling machine. The measurements were performed throughout one reactor maintenance cycle. The resultant annual cumulative dose of gamma/X-rays on the two spots of the fuel handling machine were 18.47 Mrad and 76.50 Mrad, and those of the neutrons were 17.51 krad and 60.67 krad. The measured radiation level is high enough to degrade certain cable insulation materials that may result in electrical insulation failure.

• Journal of Astronomy and Space Sciences
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• v.30 no.2
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• pp.107-112
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• 2013

Tissue equivalent proportional counter (TEPC) can measure the Linear Energy Transfer (LET) spectrum and calculate the equivalent dose for the complicated radiation field in space. In this paper, we developed and characterized a TEPC for radiation monitoring in International Space Station (ISS). The prototype TEPC which can simulate a 2 ${\mu}m$ of the site diameter for micro-dosimetry has been tested with a standard alpha source ($^{241}Am$, 5.5 MeV). Also, the calibration of the TEPC was performed by the $^{252}Cf$ neutron standard source in Korea Research Institute of Standards and Science (KRISS). The determined calibration factor was $k_f=3.59{\times}10^{-7}$ mSv/R.