• 한국의학물리학회지:의학물리
• /
• 제29권4호
• /
• pp.101-105
• /
• 2018

In this paper, a neutron moderation system for boron neutron capture therapy (BNCT) based on a $^{252}Cf$ neutron source is proposed. Different materials have been studied in order to produce a high percentage of epithermal neutrons. A moderator with a construction mixture of $AlF_3$ and Al, three reflectors of $Al_2O_3$, BeO, graphite, and seven filters (Bi, Cu, Fe, Pb, Ti, a two-layer filter of Ti+Bi, and a two-layer filter of Ti+Pb) is considered. The MCNPX simulation code has been used to calculate the neutron and gamma flux at the output window of the neutronic system. The results show that the epithermal neutron flux is relatively high for four filters: Ti+Pb, Ti+Bi, Bi, and Ti. However, a layer of Ti cannot reduce the contribution of ${\gamma}$-rays at the output window. Although the neutron spectra filtered by the Ti+Bi and Ti+Pb overlap, a large fraction of neutrons (74.95%) has epithermal energy when the Ti+Pb is used as a filter. However, the percentages of the fast and thermal neutrons are 25% and 0.5%, respectively. The Bi layer provides a relatively low epithermal neutron flux. Moreover, an assembly configuration of 30% $AlF_3+70%$ Al moderator/$Al_2O_3$ reflector/a two-layer filter of Ti+Pb reduces the fast neutron flux at the output port much more than other assembly combinations. In comparison with a recent model suggested by Ghassoun et al., the proposed neutron moderation system provides a higher epithermal flux with a relatively low contamination of gamma rays.

• Journal of Radiation Protection and Research
• /
• 제19권1호
• /
• pp.13-22
• /
• 1994

CANDU 6 중수형 원자로 운전중에 Calandria Shell내에서 발생하는 $(n,\;{\gamma})$ 반응유발 열중성자속분포와 CANDU 6 발전소의 측면 및 하단 차폐구조에서의 방사선 선량률을 계산하기 위하여 몬테칼로 방법을 이용한 MCNP 4.2 코드를 사용하였다. 계산결과, Mainshell, Annular Plate와 Subshell내 의 열중성자속분포는 $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$로 나타났고, 이는 DOT 4.2 코드의 계산결과와 비교해 볼 때 약간 큰 값들의 분포를 보여주고 있다. 이 계산결과의 응용으로서 작업자 접근가능지역 (Worker Accessible Areas)에서의 감마선량률을 계산해본 결과 설계목표치인 $6{\mu}Sv/h$보다 낮은 값을 주는 것으로 나타났다. $(n,\;{\gamma})$ 반응유발 열중성자속분포에 대한 MCNP 4.2 코드의 계산결과는 CANDU 6형 원자로의 방사선 차폐해석에 중요한 자료로 널리 이용될 수 있을 것이다.

• 한국방사선학회논문지
• /
• 제10권5호
• /
• pp.337-341
• /
• 2016

중성자에너지 영역 0.003 eV에서 10 eV에 대해 천연 Sm의 Sm(n,${\gamma}$) 반응에 대한 중성자 포획단면적을 측정하였다. 교토대학교 원자로실험소의 46-MeV 전자선형가속기에서 발생되는 전자의 광핵반응에 의한 중성자를 사용하였고 TOF 방법으로 측정하였다. 사용한 검출기는 12개의 BGO($Bi_4Ge_3O_{12}$) 섬광체로 구성되었고 이 검출장치로 Sm(n,${\gamma}$) 반응으로부터 나오는 즉발감마선을 측정하였다. 검출장치는 중성자 생성 위치로부터 $12.7{\pm}0.02m$ 위치에 설치되었으며 $^{10}B(n,{\alpha}{\gamma})^7Li$ 반응을 이용해 Sm 시료에 입사되는 중성자 선속을 구하였다. 또한 중성자 선속의 변화를 확인하기 위해 $BF_3$ 검출기로 모니터링 하였다. Sm(n,${\gamma}$) 반응단면적 측정결과는 BROND 2.2에 의한 평가결과와 J. C. Chou 및 V. N. Kononov 의 측정값과 비교하였다.

• Journal of Radiation Protection and Research
• /
• 제6권1호
• /
• pp.8-24
• /
• 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.

• Nuclear Engineering and Technology
• /
• 제34권1호
• /
• pp.30-41
• /
• 2002

The neutron beam is fully characterized for the prompt gamma activation analysis facility at Hanaro in the Korea Atomic Energy Research Institute(KAERI). The facility uses thermal neutrons which are diffracted vertically from a horizontal beam port by a set of pyrolytic graphite(PG) crystals positioned at the Bragg angle of 45" Neutron spectra, neutron flux and Cd-ratio are determined for the three extraction modes of diffracted beam by means of the theoretical and experimental efforts. To obtain theoretical result, the reflectivity of pyrolytic graphite is calculated in the diffraction model for mosaic crystal and the angular divergence after diffraction by mosaic crystal is estimated from Monte Carlo simulation. The time-of-flight spectrometer and gold activation wire are used for measuring the neutron spectra. Both the calculated and measured spectra have proven that the unique feature of polychromatic beam obtained by PG crystals are useful for PGAA. The thermal neutron flux of 7.9$\times$107 n/cm$^2$s and the Cd-ratio of 266 for gold have been achieved at the sample position while the reactor operates at 24 MW The uniformity of beam flux is 12% in the central 1$\times$1 cm$^2$ area. Finally, the beam is briefly characterized by the effective velocity and temperature which are determined by measuring the prompt Y-ray spectra for thin and thick boron samples.ples.

• Nuclear Engineering and Technology
• /
• 제49권4호
• /
• pp.787-791
• /
• 2017

Americium-beryllium (Am-Be; n, ${\gamma}$) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

• Nuclear Engineering and Technology
• /
• 제12권3호
• /
• pp.171-179
• /
• 1980

중성자 및 감마선에 대한 선량율 환산인자(flux-to-dose-rate conversion factors)를 최대흡수선량 개념을 근거로 하여 계산하였다. 중성자 및 감자선에 대한 선량율 군산인자는 에너지 범위가 각각 2.5$\times$$10^{-8}$ 20MeV 및 0.01-15MeV에 대하여 계산하였다. 이제까지 선량율 환산인자는 단일에너지에 대한 값이 였었는데 본 연구에서는 유사인체조직 (phantom)내에서 방사선의 에너지 분포가 직선적이 아니라고 가정하여 계산되었다. 특히 DLC-23, DLC-27, DLC-31 등 핵정수 자료의 각 근에 적합한 선량율 환산인자를 결정하였다는 점이 특색이다. 결과적으로 ANSI N666에 있는 값과 본 연구에서 계산된 값이 잘 일치된다는 것을 확인하였고, 본 결과는 어떤 방사선장에서도 중성자나 감마선 선량율 분포를 계산하는데 이용될 수 있고, 방사선 차폐해석, 방사선방어, radiation dosimetry 등에 필요한 값이 될 것이다.

• Nuclear Engineering and Technology
• /
• 제55권12호
• /
• pp.4329-4334
• /
• 2023

In this research, a prompt gamma neutron activation analysis (PGNAA) system is designed and constructed based on the use of a low power research reactor. For this purpose, despite the fact that this reactor did not include beam tubes, a thermal neutron beam line is installed inside the reactor tank. The extraction of the beam line from inside the tank made it possible to provide the neutron flux from the order of 106 n.cm^-2.s^-1. Also, because the beam line is installed in a tangential position to the reactor core, its gamma level has been minimized. Also, a suitable radiation shield is considered for the detector to minimize the background radiation and prevent radiation damage to the detector. Calculations and measurements are done in order to characterize this system, as well as spectrometry of several samples. The results of evaluations and experiments show that this system is suitable for performing PGNAA.

• Journal of Radiation Protection and Research
• /
• 제41권4호
• /
• pp.378-383
• /
• 2016

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.

• 분석과학
• /
• 제16권5호
• /
• pp.426-429
• /
• 2003

Neutron induced prompt gamma-ray spectroscopy (NIPS) system equipped with a $^{252}Cf$ neutron source and a n-type coaxial HPGe detector was installed for the quantitative analysis of aqueous samples in KAERI, Korea. Since the thermal neutron flux for the $^{252}Cf$ neutron source is relatively low compared to that for the reactor, the use of a thermal neutron reflector in the NIPS system may lead to improved results. The enhancement by using various reflectors was carried out by comparing the Cl peak with or without a cadmium plate between sample and the $^{252}Cf$ source. The use of pyrolitic graphite as a reflector provided a good result.