• The Journal of Korean Society for Radiation Therapy
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• v.17 no.2
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• pp.141-145
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• 2005

Purpose : It is known that the neutron is generally generated from the photon, its energy is larger than 10 MV. The neutron is leaked in the container inspection system installed at the customs though its energy is below 9 MV. It is needed that the spacial effect of the neutrons released from radiation treatment machine, linac, installed in the medical canter. Materials and Methods : The medical linear accelerator(Clinac 1800, varian, USA) was used in the experiment. Measuring neutron was used bubble detector(Bubble detector, BDPND type, BTI, Canada) which was created bubble by neutron. The bubble detector is located on the medical linear accelerator outskirt in three different distance, 30, 50, 120 cm and upper, lower four point from the iso-center. In addition, for effect on protect material we have measured eight points which are 50 cm distance from iso-center. The SAD(source-axis-distance), distance from photon source to iso-center, is adjusted to 100 cm and the field size is adjusted to $15{\times}15cm^2$. Irradiate 20 MU and calculate the dose rate in mrem/MU by measuring the number of bubble. Results : The neutron is more detected at 5 position in 30, 50 cm, 7 position in 120 cm and with wedge, and 2 position without mount. Conclusion : Though detection position is laid in the same distance in neutron measurement, the different value is shown in measuring results. Also, neutron dose is affected by the additional structure, the different value is obtained in each measurement positions. So, it is needed to measure and evaluate the neutron dose in the whole space considering the effect of the distance, angular distribution and additional structure.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.220-224
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• 1997

하나로에 설치할 냉중성자원은 물리, 화학 및 재료과학 분야에 폭넓게 활용되는 기반 장치이며 4 $\AA$ 이상의 중성자 파장에서 높은 중성자속을 얻기 위해서는 감속재의 선택이 중요하다. 이 보고에서는 감속재로 액체 수소와 액체 중수소를 사용하는 경우를 비교하였다. 계산은 몬테칼로 코드인 MCNP를 이용하고 액체 수소와 액체 중수소에 대한 산란법칙을 적용했다. Semi-analytic 방법과 MCNP 해석을 통해 중성자온도와 이득을 계산하였으며 전체적으로는 Semi-analytic 방법과 MCNP 해석이 근접함을 확인하였다.

• Journal of the Korean Vacuum Society
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• v.14 no.3
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• pp.126-131
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• 2005

Small-angle neutron scattering (SANS) is a very powerful techniques for nanostructure characterization. In this paper, we report the conceptual design and technical description of the 40 in small angle neutron scattering (SANS) insoument which is being developed by KAIST and KAERI for installation at 30 MW HANARO research reactor. For the optimal design of 40 m SANS, a series of computer simulation were pe.to.mod. The Q-.ange of the 40m SANS inst.ument is $0.0005\;\AA^{-1}-1.0\AA^{-1}$ which is a world top-class SANS Q-range. When the cold neutron spectrum used in the simulation is realized, the neutron flux at sample position is expected to be comparable to the current state-of-art SANS instrument in the world.

• Proceedings of the Materials Research Society of Korea Conference
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• 2002.05a
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• pp.89-89
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• 2002

• Proceedings of the Korea Crystallographic Association Conference
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• 2004.11a
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• pp.13-13
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• 2004

• Polymer Science and Technology
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• v.6 no.4
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• pp.403-412
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• 1995

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.66-70
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• 1989

In routine testing, the radioactive neutron sources are particularly suitable for producing standard. neutron fields. The ISO TC-85 has proposed neutron reference radiation for the calibration of neutron measuring devices used for radiation protection purposes. Radiation laboratory of KSRI has installed a standard irradiation facility using $^{252}Cf$ and $^{241}Am-Be$ sources for calibrating personal dosimeters according to the recommendations given in ISO TC-85. In this study, correction factors for calibration related to neutron scattering and anisotropy are obtained by experiments with commercial rem meter for demonstration purposes.

• 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%).

• Journal of Radiation Protection and Research
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• v.1 no.1
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• pp.1-9
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• 1976

A suggestion has been made for neutron dosimetric techniques using activation and threshold detectors in criticality accidents. Neutron dosimetrical parameters, namely, the fission spectrum-averaged cross-sections of some threshold reactions and fluence-to-dose conversion factors have been calculated by the use of an electronic computer. It appears that detectors having comparatively high threshold energy give more fine information on spectral deformation in criticality accidents, while detectors with low threshold energy are of usefulness for measuring fast neutron fluence regardless of fissioning types. Unexpectedly it is found that the fission spectrum-averaged cross sections of the $^{32}S(n,\;p)^{32}P$ reaction is not sensitive to analytical forms of fission neutron spectrum: the modified Cran-berg and Maxwellian forms. In addition, the fluence-to-dose conversion factors seem to be insensitive to both spectral functions and fissioning types.

• Journal of the Korean Society of Radiology
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• v.13 no.7
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• pp.917-924
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• 2019

The passive scattering system nozzle of the proton therapy accelerator was simulated to evaluate the neutrons generated by each component in each nozzle by energy. The Monte Carlo N-Particle code was used to implement spread out Bragg peak with proton energy 220 MeV, reach 20 cm, and 6 cm length used in the treatment environment. Among the proton accelerator components, neutrons were the highest in scatterers, and the neutron flux decreased as it moved away from the central flux of the proton. This study can be used as a basic data for the evaluation of the radiation necessary for the maintenance and dismantling of proton accelerators.