• Title, Summary, Keyword: PHITS

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Beam Dump Fabrication and Simulation by Using PHITS Code and an ANSYS Analysis

  • Kim, Seong Jun;Park, Jin Yong;Lee, Byoung-Seob;Hong, Jonggi;Won, Mi-Sook;Ok, Jung-Woo;Choi, Seyong
    • Journal of the Korean Physical Society
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    • v.73 no.8
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    • pp.1115-1117
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    • 2018
  • The design of a beam dump for the energy range less than 20 MeV is reported. The beam dump was developed for the beam extraction commissioning stage of the Korea Basic Science Institute's(KBSI's) heavy ion facility. At present, the downstream section design of the diagnostic chamber of the beam line was completed, as has the fabrication of a radio-frequency quadrupole(RFQ) section. We are now performing frequency tuning for optimization. For reduction and elimination of the generated beam, the beam dump will be located downstream of the diagnostic beam line, and the beam dump was designed to withstand a heavy-ion beam energy of less than 20 MeV. The PHITS code and ANSYS steady-state thermal analysis were adapted for simulating the heat transfer and determining the numbers and types of secondary particles produced by bombardment of the beam's dump material, respectively.

Measurement of Gamma-ray Yield from Thick Carbon Target Irradiated by 5 and 9 MeV Deuterons

  • Araki, Shouhei;Kondo, Kazuhiro;Kin, Tadahiro;Watanabe, Yukinobu;Shigyo, Nobuhiro;Sagara, Kenshi
    • Journal of Radiation Protection and Research
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    • v.42 no.1
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    • pp.16-20
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    • 2017
  • Background: The design of deuteron accelerator neutron source facilities requires reliable yield estimation of gamma-rays as well as neutrons from deuteron-induced reactions. We have so foar measured systematically double-differential thick target neutron yields (DDTTNYs) for carbon, aluminum, titanium, copper, niobium, and SUS304 targets. In the neutron data analysis, the events of gamma-rays taken simultaneously were treated as backgrounds. In the present work, we have re-analyzed the experimental data for a thick carbon target with particular attention to gamma-ray events. Materials and Methods: Double-differential thick target gamma-ray yields from carbon irradiated by 5 and 9 MeV deuterons were measured using an NE213 liquid organic scintillator at the Kyushu University Tandem accelerator Laboratory. The gamma-ray energy spectra were obtained by an unfolding method using FORIST code. The response functions of the NE213 detector were calculated by EGS5 incorporated in PHITS code. Results and Discussion: The measured gamma-ray spectra show some pronounced peaks corresponding to gamma-ray transitions between discrete levels in residual nuclei, and the measured angular distributions are almost isotropic for both the incident energies. Conclusion: PHITS calculations using INCL, GEM, and EBITEM models reproduce the spectral shapes and the angular distributions generally well, although they underestimate the absolute gamma-ray yields by about 20%.

Measurement of Neutron Production Double-differential Cross-sections on Carbon Bombarded with 430 MeV/Nucleon Carbon Ions

  • Itashiki, Yutaro;Imahayashi, Youichi;Shigyo, Nobuhiro;Uozumi, Yusuke;Satoh, Daiki;Kajimoto, Tsuyoshi;Sanami, Toshiya;Koba, Yusuke;Matsufuji, Naruhiro
    • Journal of Radiation Protection and Research
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    • v.41 no.4
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    • pp.344-349
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    • 2016
  • Background: Carbon ion therapy has achieved satisfactory results. However, patients have a risk to get a secondary cancer. In order to estimate the risk, it is essential to understand particle transportation and nuclear reactions in the patient's body. The particle transport Monte Carlo simulation code is a useful tool to understand them. Since the code validation for heavy ion incident reactions is not enough, the experimental data of the elementary reaction processes are needed. Materials and Methods: We measured neutron production double-differential cross-sections (DDXs) on a carbon bombarded with 430 MeV/nucleon carbon beam at PH2 beam line of HIMAC facility in NIRS. Neutrons produced in the target were measured with NE213 liquid organic scintillators located at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. Results and Discussion: Neutron production double-differential cross-sections for carbon bombarded with 430 MeV/nucleon carbon ions were measured by the time-of-flight method with NE213 liquid organic scintillators at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. The cross sections were obtained from 1 MeV to several hundred MeV. The experimental data were compared with calculated results obtained by Monte Carlo simulation codes PHITS, Geant4, and FLUKA. Conclusion: PHITS was able to reproduce neutron production for elementary processes of carbon-carbon reaction precisely the best of three codes.

Development of Neutron Skyshine Evaluation Method for High Energy Electron Accelerator Using Monte Carlo Code (몬테카를로 코드를 이용한 고에너지 전자가속기의 중성자 skyshine 평가방법 개발)

  • Oh, Joo-Hee;Jung, Nam-Suk;Lee, Hee-Seock;Ko, Seung-Kook
    • Journal of Radiation Protection and Research
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    • v.38 no.1
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    • pp.22-28
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    • 2013
  • The skyshine effect is an essential and important phenomenon in the shielding design of the high energy accelerator. In this study, a new estimation method of neutron skyshine was proposed and was verified by comparison with existing methods. The effective dose of secondary neutrons and photons at the locations that was far away from high-energy electron accelerator was calculated using FLUKA and PHITS Monte Carlo code. The transport paths of secondary radiations to reach a long distance were classified as skyshine, direct, groundshine and multiple-shine. The contribution of each classified component to the total effective dose was evaluated. The neutrons produced from the thick copper target irradiated by 10 GeV electron beam was applied as a source term of this transport. In order to evaluate a groundshine effect, the composition of soil on the PAL-XFEL site was considered. At a relatively short distance less than 50 m from the accelerator tunnel, the direct and groundshine components mostly contributed to the total effective dose. The skyshine component was important at a long distance. The evaluated dose of neutron skyshine agreed better with the results using Rindi's formula, which was based on the experimental results at high energy electron accelerator. That also agreed with the estimated dose using the simple evaluation code, SHINE3, within about 20%. The total effective dose, including all components, was 10 times larger than the estimated doses using other methods for this comparison. The influence of multiple-shine path in this evaluation of the estimation method was investigated to be bigger than one of pure skyshine path.

Study on Concrete Activation Reduction in a PET Cyclotron Vault

  • Bakhtiari, Mahdi;Oranj, Leila Mokhtari;Jung, Nam-Suk;Lee, Arim;Lee, Hee-Seock
    • Journal of Radiation Protection and Research
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    • v.45 no.3
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    • pp.130-141
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    • 2020
  • Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd2O3, B4C, polyethylene (PE), and borated (5 wt% natB) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B4C and Gd2O3 were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1- or 0.4-cm-thick B4C or Gd2O3 behind 10 cm-thick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cm-thick Gd2O3+10 cm-thick PE or 0.1 cm-thick B4C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B4C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.

Attenuation curves of neutrons from 400 to 550 Mev/u for Ca, Kr, Sn, and U ions in concrete on a graphite target for the design of shielding for the RAON in-flight fragment facility in Korea

  • Lee, Eunjoong;Kim, Junhyeok;Kim, Giyoon;Kim, Jinhwan;Park, Kyeongjin;Cho, Gyuseong
    • Nuclear Engineering and Technology
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    • v.51 no.1
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    • pp.275-283
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    • 2019
  • Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

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.

TET2MCNP: A Conversion Program to Implement Tetrahedral-mesh Models in MCNP

  • Han, Min Cheol;Yeom, Yeon Soo;Nguyen, Thang Tat;Choi, Chansoo;Lee, Hyun Su;Kim, Chan Hyeong
    • Journal of Radiation Protection and Research
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    • v.41 no.4
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    • pp.389-394
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    • 2016
  • Background: Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET2MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. Materials and Methods: TET2MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET2MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. Results and Discussion: To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET2MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. Conclusion: In the present study, we have developed a computer program, TET2MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code.