• Title, Summary, Keyword: MCNPX

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Comparison between General X-ray Tube Modeling and Apply Energy-spectrum by MCNPX Simulation (MCNPX 시뮬레이션을 이용한 엑스선관 모델링 비교)

  • Jung, Jae-Hong;Lee, Jun-Jae;Lee, Woo-Pil;Ahn, Hyun-Jun;Kim, Sang-Hyun
    • Journal of Radiation Industry
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    • v.12 no.3
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    • pp.209-216
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    • 2018
  • The MCNPX (Monte Carlo N-Particle Extended) is a program defined as the simulation using a stochastic system for decision making in an uncertain situation. In this study, general X-ray tube modeling and apply energy-spectrum were compared physical characteristics including the photon fluence, percent depth dose (PDD) and energy-spectrum. For photon fluence, X-ray tube modeling was lower than apply energy-spectrum modeling. For PDD and energy-spectrum, X-ray tube modeling was similar to other modeling. However, simulation time at X-ray tube modeling was higher than other modeling. Consequently, apply energy-spectrum modeling could be useful in terms of simulation time with reliability in MCNPX simulation.

Shielding Calculations of Accelerator Facility for Medical Isotope Production using MCNPX Code (MCNPX 코드를 이용한 의료용 방사성동위원소 생산을 위한 가속기 시설의 방사선차폐 및 선량 계산)

  • Seo Kyu-Seok;Kim Chan-Hyeong
    • Progress in Medical Physics
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    • v.15 no.4
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    • pp.210-214
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    • 2004
  • 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.

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Calculation of Energy Spectra for 6 MeV Electron Beam of LINAC Using MCNPX (MCNPX를 이용한 선형가속기의 6 MeV 전자선에 대한 에너지분포 계산)

  • Lee, Jeong-Ok;Jeong, Dong-Hyeok
    • Progress in Medical Physics
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    • v.17 no.4
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    • pp.224-231
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    • 2006
  • The electron energy spectra for 6 MeV electron beam were calculated using a MCNPX code. The head of the linear accelerator (ML6M; Mitsubishi, Japan) was modelled for this study. The energy spectrum of the initial electron beam was assumed to be Gaussian and the mean energy was determined by evaluating the measured and calculated values of $R_{50}$ and dose profiles in air. The energy distributions for electrons and photons at the interested points in the head of the linear accelerator were calculated by appling the Initial beam parameters. The effect of contaminant photons on depth dose curves were estimated by the photon energy spectra at the end of the applicator.

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Study of Radiation dose Evaluation using Monte Carlo Simulation while Treating Extrahepatic Bile Duct Cancer with High Dose Rate Intraluminal Brachytherapy (간외 담도암 고선량률 관내근접방사선치료 시 몬테카를로 시뮬레이션을 통한 주변장기의 선량평가 연구)

  • Park, Ju-Kyeong;Lee, Seung-Hoon;Cha, Seok-Yong;Lee, Sun-Young
    • The Journal of the Korea Contents Association
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    • v.14 no.2
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    • pp.467-474
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    • 2014
  • The relative dose calculated by MCNPX and the relative dose measured by ionization chamber and solid phantoms evaluated the accuracy comparing with Monte Carlo simulation. In order to apply Monte Carlo simulation the intraluminal brachytherapy of extrahepatic bile duct cancer, 192Ir sealed radioactive source replicate, Bile duct and surrounding organs were made using KMIRD phantom based on a South Korea standard man. To check the absorbed dose of normal organs around bile duct, we set the specific effective energy and initial radioactivity to 1 Ci using MCNPX. Evaluation of the accuracy of the Monte Carlo simulation, the difference of the relative dose is the most 1.96% that satisfy the criteria that is the relative error less than 2% suggested by MCNPX code. In addition, The specific effective energy and absorbed dose of normal organs that were relatively adjacent to bile duct such as right side of kidney, liver, pancreas, transverse colon, spinal cord, stomach and small intestine were relatively high. on the contrary, the organs that were relatively distant to bile duct such as left side of kidney, spleen, ascending colon, descending colon and sigmoid colon were relatively low.

Calculation of Neutron Energy Distribution from the Components of Proton Therapy Accelerator Using MCNPX (MCNPX를 이용한 양성자 치료기의 구성품에서 발생하는 중성자 에너지 분포계산)

  • Bae, Sang-Il;Shin, Sang-Hwa
    • 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.

Evaluation of Photoneutron Dose in Radiotherapy Room Using MCNPX (MCNPX를 이용한 방사선 치료실의 광중성자 선량 평가)

  • Park, Eun-Tae
    • The Journal of the Korea Contents Association
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    • v.15 no.6
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    • pp.283-289
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    • 2015
  • Recently, high energy photon radiotherapy is a growing trend for increasing therapy results. Commonly, if you use high energy photons above 6~8 MeV nominal accelerator voltage, It lead the photo-nuclear reaction and the generation of photo-neutron are accompanied and these problematic factors are issued in the view of radiation protection. Therefore, in this study analyzed for dose distribution of photo-neutron in radiotherapy room based on MCNPX. As a result, absorbed dose is increased sharply from 10 MV to 12 MV. It was founded that the rapid increasement of photoneutron fluence was related to the absorbed dose at above 10 MV. Also, in case of the recommendation of ICRP 103, the outcome of an exchanged equivalent dose which based on calculated an absorbed dose, showed lower equivalent dose than ICRP 60 by reflecting the contribution of secondary photon for absorbed dose of human body in the low energy band.

Fabrication of Fiber-optics Detector for Measuring Radioactive Waste (방사성 오염도 측정을 위한 광섬유 검출기 제작)

  • Kim, Jeong-Ho;Joo, Koan-Sik
    • Journal of IKEEE
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    • v.19 no.3
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    • pp.282-287
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    • 2015
  • In this study, an optical fiber detector was constructed by using a Ce:GAGG scintillator, optical fiber, and photomultiplier. The single crystal size of the scintillator was set to $3{\times}3{\times}20mm^3$ after simulating the counting efficiency of gamma rays in the scintillator by using the MCNPX code. The constructed detector used the standard gamma ray sources $^{137}Cs$ and $^{133}Ba$ to measure radiation and analyze the spectral characteristics of gamma rays. The resulting trend curve showed excellent linearity with an R-squared value of 0.99741, and the detector characteristics were found to vary 2% or less with distance based on comparison with the MCNPX value. Furthermore, the spectroscopic analysis of the gamma ray energy from the single-ray and mixed-ray sources showed that $^{137}Cs$ had its peak energy at 662 keV, and $^{133}Ba$ had at 356 keV. It seems that if the fiber-optics detector is used, working hours and exposure of worker can be reduced.

Comparison of Physics Model for 600 MeV Protons and 290 MeV·n-1 Oxygen Ions on Carbon in MCNPX

  • Lee, Arim;Kim, Donghyun;Jung, Nam-Suk;Oh, Joo-Hee;Oranj, Leila Mokhtari;Lee, Hee-Seock
    • Journal of Radiation Protection and Research
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    • v.41 no.2
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    • pp.123-131
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    • 2016
  • Background: With the increase in the number of particle accelerator facilities under either operation or construction, the accurate calculation using Monte Carlo codes become more important in the shielding design and radiation safety evaluation of accelerator facilities. Materials and Methods: The calculations with different physics models were applied in both of cases: using only physics model and using the mix and match method of MCNPX code. The issued conditions were the interactions of 600 MeV proton and $290MeV{\cdot}n^{-1}$ oxygen with a carbon target. Both of cross-section libraries, JENDL High Energy File 2007 (JENDL/HE-2007) and LA150, were tested in this calculation. In the case of oxygen ion interactions, the calculation results using LAQGSM physics model and JENDL/HE-2007 library were compared with D. Satoh's experimental data. Other Monte Carlo calculations using PHITS and FLUKA codes were also carried out for further benchmarking study. Results and Discussion: It was clearly found that the physics models, especially intra-nuclear cascade model, gave a great effect to determine proton-induced secondary neutron spectrum in MCNPX code. The variety of physics models related to heavy ion interactions did not make big difference on the secondary particle productions. Conclusion: The variations of secondary neutron spectra and particle transports depending on various physics models in MCNPX code were studied and the result of this study can be used for the shielding design and radiation safety evaluation.

An extensive investigation on gamma ray shielding features of Pd/Ag-based alloys

  • Agar, O.;Sayyed, M.I.;Akman, F.;Tekin, H.O.;Kacal, M.R.
    • Nuclear Engineering and Technology
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    • v.51 no.3
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    • pp.853-859
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    • 2019
  • A comprehensive study of photon interaction features has been made for some alloys containing Pd and Ag content to evaluate its possible use as alternative gamma radiations shielding material. The mass attenuation coefficient (${\mu}/{\rho}$) of the present alloys was measured at various photon energies between 81 keV-1333 keV utilizing HPGe detector. The measured ${\mu}/{\rho}$ values were compared to those of theoretical and computational (MCNPX code) results. The results exhibited that the ${\mu}/{\rho}$ values of the studied alloys are in the same line with results of WinXCOM software and MCNPX code results at all energies. Moreover, Pd75/Ag25 alloy sample has the maximum radiation protection efficiency (about 53% at 81 keV) and lowest half value layer, which shows that Pd75/Ag25 has superior gamma radiation shielding performance among the other compared alloys.