Earticle

본 연구에서는 미래 파이로 시설에서의 핵물질 계량 연구를 위하여 개발하고 있는 통합비파괴측정장치(Unified Non-Destructive Assay, UNDA)의 중성자 검출 효율을 MCNPX 코드를 이용하여 평가하였다. 검출 효율 평가는 두 개의 다른 설계안의 UNDA에 대하여 수행되었으며, 252Cf 중성자 발생 선원 위치에 따른 검출 효율 평가와 감손우라늄 의 용기 두께 및 위치에 따른 검출 효율 평가를 수행하였다. 252Cf 중성자 선원의 위치에 따른 UNDA의 검출 효율 결과는 6.83%부터 13.35%까지 분포로 나타났으며, 252Cf 선원이 장치 내부의 상단에 위치할수록 검출 효율은 증가 후 감소하는 경향을 나타냈고, 선원이 외각에 위치될수록 효율이 증가하는 경향을 보였다. 감손우라늄 용기의 두께 및 위치에 따른 검출 효율 평가에서는 용기 두께가 증가할수록 검출 효율은 낮아지는 경향을 보이며, 용기 위치가 장치 상부에 위치될수록 효율은 감소하고, 외각에 위치할수록 효율은 증가하였다. 검출 효율은 252Cf 선원의 경우보다 약간 높게 나타났다(10.31~13.61%). 또한, 장치 상단에 고밀도 폴리에틸렌 덮개가 있는 설계안이 덮개가 없는 설계안 보다 평균적으로 약 2% 정도 중성자 검출 효율이 높은 것으로 평가되었다.

In this study, neutron detection efficiency of the UNDA system, which has been developed for study on nuclear material accountancy in a future pyro-process facility, was evaluated by using the MCNPX code. The detection efficiency was evaluated as a function of (1) positions of 252Cf neutron source in the axial and radial directions, and (2) thicknesses and locations of the container filled with the depleted uranium materials for two different designs of the UNDA. In the case of 252Cf source positions, detection efficiency was distributed from 6.83% to 13.35%. As 252Cf source was positioned at upper part in the axial direction, detection efficiency was decreased after a slight increase. On the other hands, as 252Cf source was positioned at outer part in the radial direction, detection efficiency was increased. In the case of container thickness, there was a slight decline when the thickness was increased. As the container was located at upper part, detection efficiency was decreased and as the container was located at outer part, detection efficiency was increased. Detection efficiency was varied from 10.31% to 13.61%. These values were higher than that of 252Cf source case. The UNDA with polyethylene cover has about 2% higher detection efficiency than the UNDA without the cover.

Background: Shielding properties of compound or mixture is presented in terms of mass attenuation coefficients using Monte Carlo simulation. Mass attenuation coefficients of cement, gypsum and the mixture of gypsum and PbCO3 has been investigated using monte carlo MCNPX. Materials and Methods: The mass attenuation coefficients of cement, gypsum and the mixture of gypsum and PbCO3 were calculated for photon energies 365.5, 661.6, 1,173.2, and 1,332.5 keV energies. Results and Discussion: The simulated values of mass attenuation coefficients were compared avaialable experimental results, theoretical values by XCOM and found good comparability of the results. Conclusion: Standard simulation geometry used in the present investigation would be very useful for various types of sample for shielding and dosimetry applications.

PET에 사용되는 조영제는 생산과정 중에 다량의 중성자가 발생한다. 발생된 중성자는 주로 콘크리트 구조물로 차폐를 하게 되며 가속기 시설의 차폐 평가는 구조물 외부로 방출되는 방사선의 선량을 측정하게 된다. 즉 콘크리트를 통과하면서 에너지를 잃은 중성자와 콘크리트를 이루는 물질과 중성자간의 상호작용으로 생성되는 광자의 선량을 측정하여 선량을 평가하게 된다. MCNPX 코드2)를 이용하여 가속기 시설의 콘크리트 구조물 외부로 방출되는 중성자 선량과 광자선량을 계산한 결과, 원자력법에서 정한 법정 제한 선량에 훨씬 못 미치는 것을 알 수 있었다.

Since production of radioactive isotope for using PET, a lot of neutrons were produced. The produced neutrons were mainly shielded by concrete facility. Secondary photons are generated and emitted from the concrete shielding wall of the PET cyclotron since the proton-generated neutrons are thermalized and absorbed in the concrete wall and emit secondary radiations, i.e., photons. This study calculated neutron dose and photon dose at outside of the accelerator facility using MCNPX code. As results of the calculation, total dose were calculated less than limited dose by law.

갑상선암 진단 방사성의약품인 $^{123}I$ 생산을 목적으로 한 가스타겟챔버를 개발하고 MCNPX를 이용해 30MeV 빔에너지 가 가스타겟챔버에 어떻게 들어가는지와 들어갔을 경우의 $^{124}Xe$와 핵반응은 어떻게 발생하는지를 모델링하였다. 빔에너지 가 확산되어 가스타겟챔버 내경에 맞아 에너지 손실이 생긴다. 그것은 즉 손실된 에너지가 열로 바뀜으로 타겟챔버가 변형이 일어나기 않게 냉각수를 이용한다. 쿨링시스템도 타겟챔버를 효율적으로 냉각하기위해 냉각수라인을 나선형으로 설계하였다. KIRAMS에서 보유하고 있는 사이크로트론 C30을 이용하여 30MeV 에너지에 100A 빔을 조사해 $^{124}Xe(p,2n)$, $^{124}Xe(p,n)$, $^{124}Xe(p,pn)$ 각각의 핵반응이 일어나는걸 알 수 있었고 생산량을 예측 할 수 있었다.

Gas target chamber has been developed for producing $^{123}I$ which is radiopharmaceuticals for diagnosis of thyroid cancer, and modeled how to occur nuclear reaction between chamber and $^{124}Xe$ with energy 30MeV inside the gas target chamber by using the MCNPX. The beam energy was lost as the beam spread when beam hit inside the gas target chamber. The cooling water was used not to change the gas target chamber as loss of energy transfer to the thermal energy. Spiral cooling line was designed for cooling the target chamber efficiently. By using the c30 cyclotron, $^{124}Xe(p,2n)$, $^{124}Xe(p,n)$, $^{124}Xe(p,pn)$ nuclear reactions were studied. In this study, we predict the production yield.