Journal articles: 'Monte Carlo N-Particle Transport Code'

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Romano, Paul K., and Benoit Forget. "The OpenMC Monte Carlo particle transport code." Annals of Nuclear Energy 51 (January 2013): 274–81. http://dx.doi.org/10.1016/j.anucene.2012.06.040.

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Li Gang, 李刚, 张宝印 Zhang Baoyin, 邓力 Deng Li, 胡泽华 Hu Zehua, and 马彦 Ma Yan. "Development of Monte Carlo particle transport code JMCT." High Power Laser and Particle Beams 25, no. 1 (2013): 158–62. http://dx.doi.org/10.3788/hplpb20132501.0158.

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Žohar, Andrej, Žiga Štancar, Paola Batistoni, Sean Conroy, Luka Snoj, and Igor Lengar. "VALIDATION OF SERPENT FOR FUSION NEUTRONICS ANALYSIS AT JET." EPJ Web of Conferences 247 (2021): 18001. http://dx.doi.org/10.1051/epjconf/202124718001.

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Fusion neutronics analysis before and after experiments at JET is traditionally performed using Monte Carlo particle transport code Monte Carlo N-Particle. For redundancy and diversity reasons there is a need of an additional Monte Carlo code, such as Serpent 2, capable of fusion neutronics analysis. In order to validate the Serpent code for fusion applications a detailed model of JET was used. Neutron fluxes and reaction rates were calculated and compared for positions outside the tokamak vacuum vessel, in the vacuum vessel above the plasma and next to a limiter inside the vacuum vessel. For all detector positions with DD and DT neutron sources the difference between neutron fluxes calculated with both Monte Carlo codes were within 2σ statistical uncertainty and for most positions (more than 90 % of all studied positions) even within 1σ uncertainty. Fusion neutronics analysis in the JET tokamak with Serpent took on average 10 % longer but this can be improved by changing the threshold value for determination of the transport method used. With the work presented in this paper the Serpent Monte Carlo code was validated to be a viable alternative to MCNP for fusion neutronics analysis for the JET tokamak.

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Peralta, Luis, and Alina Louro. "AlfaMC: A fast alpha particle transport Monte Carlo code." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 737 (February 2014): 163–69. http://dx.doi.org/10.1016/j.nima.2013.11.026.

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Hadjidoukas, P., C. Bousis, and D. Emfietzoglou. "Parallelization of a Monte Carlo particle transport simulation code." Computer Physics Communications 181, no. 5 (May 2010): 928–36. http://dx.doi.org/10.1016/j.cpc.2010.01.005.

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Punjabi, Alkesh, Allen Boozer, Maria Lam, Myung-Hee Kim, and Kathy Burke. "Monte Carlo calculations for transport due to MHD modes." Journal of Plasma Physics 44, no. 3 (December 1990): 405–30. http://dx.doi.org/10.1017/s0022377800015282.

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The three basic mechanisms that produce either classical or anomalous transport are spatial variation of magnetic field strength, spatial variation of electrostatic potential in magnetic surfaces, and loss of magnetic surfaces. A Monte Carlo code is written to study transport due to these three mechanisms interacting with collisional effects. The equations of motion are obtained from the canonical drift Hamiltonian, but non-canonical co-ordinates are used to simplify the integrations. The code is applied to the reversed-field-pinch ZT-40 and the Tokapole II. For ZT-40 the Bessel-function model is used to represent the magnetic field geometry. The effects of pitch-angle scattering, loop voltage and the break-up of magnetic surfaces resulting from resistive MHD perturbations on the drift particle trajectories are illustrated. The particle diffusion coefficients are obtained for varying amplitudes of resistive MHD perturbations. For Tokapole II the spectrum of both the ideal and resistive MHD perturbations is constructed from the experimental data. The drift trajectories for trapped and passing electrons in the presence of such perturbations are obtained. The particle diffusion coefficients for the neo-classical regime in Tokapole II are obtained for varying collision frequency. By comparing the transport coefficients for various groups of particles with the experimental data, we hope to obtain far more information on the transport mechanisms than can be obtained by the standard confinement time measurements. The various groups of particles that can be studied using the code include runaway electrons, thermal electrons, and both passing and trapped diagnostic beam ions.

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Romano, Paul K., Colin J. Josey, Andrew E. Johnson, and Jingang Liang. "Depletion capabilities in the OpenMC Monte Carlo particle transport code." Annals of Nuclear Energy 152 (March 2021): 107989. http://dx.doi.org/10.1016/j.anucene.2020.107989.

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Ilic, Radovan. "Proton therapy Monte Carlo SRNA-VOX code." Nuclear Technology and Radiation Protection 27, no. 4 (2012): 355–67. http://dx.doi.org/10.2298/ntrp1204355i.

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The most powerful feature of the Monte Carlo method is the possibility of simulating all individual particle interactions in three dimensions and performing numerical experiments with a preset error. These facts were the motivation behind the development of a general-purpose Monte Carlo SRNA program for proton transport simulation in technical systems described by standard geometrical forms (plane, sphere, cone, cylinder, cube). Some of the possible applications of the SRNA program are: (a) a general code for proton transport modeling, (b) design of accelerator-driven systems, (c) simulation of proton scattering and degrading shapes and composition, (d) research on proton detectors; and (e) radiation protection at accelerator installations. This wide range of possible applications of the program demands the development of various versions of SRNA-VOX codes for proton transport modeling in voxelized geometries and has, finally, resulted in the ISTAR package for the calculation of deposited energy distribution in patients on the basis of CT data in radiotherapy. All of the said codes are capable of using 3-D proton sources with an arbitrary energy spectrum in an interval of 100 keV to 250 MeV.

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Bernat, Robert, Luka Bakrač, Vladimir Radulović, Luka Snoj, Takahiro Makino, Takeshi Ohshima, Željko Pastuović, and Ivana Capan. "4H-SiC Schottky Barrier Diodes for Efficient Thermal Neutron Detection." Materials 14, no. 17 (September 6, 2021): 5105. http://dx.doi.org/10.3390/ma14175105.

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In this work, we present the improved efficiency of 4H-SiC Schottky barrier diodes-based detectors equipped with the thermal neutron converters. This is achieved by optimizing the thermal neutron converter thicknesses. Simulations of the optimal thickness of thermal neutron converters have been performed using two Monte Carlo codes (Monte Carlo N–Particle Transport Code and Stopping and Range of Ions in Matter). We have used 6LiF and 10B4C for the thermal neutron converter material. We have achieved the thermal neutron efficiency of 4.67% and 2.24% with 6LiF and 10B4C thermal neutron converters, respectively.

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Kowalski, Mikolaj Adam, Paul Cosgrove, Jakob Broman, and Eugene Shwageraus. "SCONE: A Student-Oriented Modifiable Monte Carlo Particle Transport Framework." Journal of Nuclear Engineering 2, no. 1 (March 8, 2021): 57–64. http://dx.doi.org/10.3390/jne2010006.

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Over the last decade, the importance of the Monte Carlo as a neutron transport calculation method has greatly increased. This paper describes a Monte Carlo particle transport framework SCONE, which aims to provide with easy-to-learn environment for graduate students to learn about Monte Carlo methods and explore new ideas. The paper lists the steps taken to enhance new user experience of SCONE and briefly discuses how the architecture supports its goals. The current version of the code is compared against Serpent and shown to provide with sufficient accuracy to be used for teaching and proof-of-concept applications.

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Shangguan Danhua, 上官丹骅, 李刚 Li Gang, 张宝印 Zhang Baoyin, and 邓力 Deng Li. "Design of sampling tools for Monte Carlo particle transport code JMCT." High Power Laser and Particle Beams 24, no. 12 (2012): 2955–58. http://dx.doi.org/10.3788/hplpb20122412.2955.

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Nguyen, T. S., and G. B. Wilkin. "Monte Carlo Calculations Applied to NRU Reactor and Radiation Physics Analyses." AECL Nuclear Review 1, no. 2 (December 1, 2012): 47–50. http://dx.doi.org/10.12943/anr.2012.00018.

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The statistical MCNP (Monte Carlo N-Particle) code has been satisfactorily used for reactor and radiation physics calculations to support NRU operation and analysis. MCNP enables 3D modeling of the reactor and its components in great detail, the transport calculation of photons (in addition to neutrons), and the capability to model all locations in space, which are beyond the capabilities of the deterministic neutronics methods used for NRU. While the simple single-cell model is efficient for local analysis in any site of NRU, the complex full-reactor model is required for calculations of the core physics and beyond-the-core radiation. By supplementing, adjusting or benchmarking the results from the existing NRU codes, the MCNP calculations provide greater confidence that NRU remains within the licence envelope.

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Deliang, Yu, Yan Longwen, Zhong Guangwu, Lu Jie, and Yi Ping. "Neutral Particle Transport in Cylindrical Plasma Simulated by a Monte Carlo Code." Plasma Science and Technology 9, no. 2 (April 2007): 133–38. http://dx.doi.org/10.1088/1009-0630/9/2/03.

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Taasti, Vicki Trier, Helge Knudsen, Michael H. Holzscheiter, Nikolai Sobolevsky, Bjarne Thomsen, and Niels Bassler. "Antiproton annihilation physics in the Monte Carlo particle transport code SHIELD-HIT12A." Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms 347 (March 2015): 65–71. http://dx.doi.org/10.1016/j.nimb.2015.02.002.

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IWASE, Hiroshi, Koji NIITA, and Takashi NAKAMURA. "Development of General-Purpose Particle and Heavy Ion Transport Monte Carlo Code." Journal of Nuclear Science and Technology 39, no. 11 (November 2002): 1142–51. http://dx.doi.org/10.1080/18811248.2002.9715305.

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Boyd, William, Adam Nelson, Paul K. Romano, Samuel Shaner, Benoit Forget, and Kord Smith. "Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code." Nuclear Technology 205, no. 7 (March 9, 2019): 928–44. http://dx.doi.org/10.1080/00295450.2019.1571828.

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Brice, D. K. "Meastri: A hybrid analytical/monte carlo code for particle transport in solids." Journal of Nuclear Materials 162-164 (April 1989): 985–89. http://dx.doi.org/10.1016/0022-3115(89)90397-8.

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Klaß, Larissa, Philipp Ritz, Marius Hirsch, John Kettler, Andreas Havenith, Andreas Wilden, and Giuseppe Modolo. "Gamma-spectrometric measurement procedure for a clearance concept of radioactively contaminated mercury from nuclear facilities." Journal of Radioanalytical and Nuclear Chemistry 329, no. 2 (June 24, 2021): 565–80. http://dx.doi.org/10.1007/s10967-021-07840-7.

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AbstractRadioactive mercury waste constitutes a significant challenge, as no approved disposal concept yet exists for such waste in Germany. This work describes a decontamination and measurement procedure for a possible clearance of mercury from nuclear facilities and release into reuse or conventional hazardous waste disposal to reduce the amount of mercury in a nuclear repository. The measurement setup and procedure were developed and evaluated including Monte-Carlo N-Particle® Transport Code (MCNP® and Monte Carlo N-Particle® are registered trademarks owned by Los Alamos National Security, LLC, manager and operator of Los Alamos National Laboratory, (Werner 2018, Werner 2017)), simulations to ensure conservative assumptions during the measurements. Results from decontaminated mercury samples show that a clearance pursuant to the German regulations would be feasible.

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Newpower, Mark, Jan Schuemann, Radhe Mohan, Harald Paganetti, and Uwe Titt. "Comparing 2 Monte Carlo Systems in Use for Proton Therapy Research." International Journal of Particle Therapy 6, no. 1 (May 3, 2019): 18–27. http://dx.doi.org/10.14338/ijpt-18-00043.1.

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Abstract Purpose: Several Monte Carlo transport codes are available for medical physics users. To ensure confidence in the accuracy of the codes, they must be continually cross-validated. This study provides comparisons between MC2 and Tool for Particle Simulation (TOPAS) simulations, that is, between medical physics applications for Monte Carlo N-Particle Transport Code (MCNPX) and Geant4. Materials and Methods: Monte Carlo simulations were repeated with 2 wrapper codes: TOPAS (based on Geant4) and MC2 (based on MCNPX). Simulations increased in geometrical complexity from a monoenergetic beam incident on a water phantom, to a monoenergetic beam incident on a water phantom with a bone or tissue slab at various depths, to a spread-out Bragg peak incident on a voxelized computed tomography (CT) geometry. The CT geometry cases consisted of head and neck tissue and lung tissue. The results of the simulations were compared with one another through dose or energy deposition profiles, r90 calculations, and γ-analyses. Results: Both codes gave very similar results with monoenergetic beams incident on a water phantom. Systematic differences were observed between MC2 and TOPAS simulations when using a lung or bone slab in a water phantom, particularly in the r90 values, where TOPAS consistently calculated r90 to be deeper by about 0.4%. When comparing the performance of the 2 codes in a CT geometry, the results were still very similar, exemplified by a 3-dimensional γ-analysis pass rate > 95% at the 2%–2-mm criterion for tissues from both head and neck and lung. Conclusion: Differences between TOPAS and MC2 were minor and were not considered clinically relevant.

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HIGUCHI, Kenji, Hiroshi TAKEMIYA, and Takuji KAWASAKI. "Load Balancing in Highly Parallel Processing of Monte Carlo Code for Particle Transport." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 40, no. 10 (1998): 798–808. http://dx.doi.org/10.3327/jaesj.40.798.

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Lee, Hyunsuk, Wonkyeong Kim, Peng Zhang, Matthieu Lemaire, Azamat Khassenov, Jiankai Yu, Yunki Jo, Jinsu Park, and Deokjung Lee. "MCS – A Monte Carlo particle transport code for large-scale power reactor analysis." Annals of Nuclear Energy 139 (May 2020): 107276. http://dx.doi.org/10.1016/j.anucene.2019.107276.

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PAVLOVIČ, MÁRIUS, KATARÍNA SEDLAČKOVÁ, ANDREA ŠAGÁTOVÁ, and IVAN STRAŠÍK. "APPLICATION OF THE S3M AND MCNPX CODES IN PARTICLE DETECTOR DEVELOPMENT." International Journal of Modern Physics: Conference Series 27 (January 2014): 1460153. http://dx.doi.org/10.1142/s2010194514601537.

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Semiconductor detectors can be used to detect neutrons if they are covered by a conversion layer. Some neutrons transfer their kinetic energy to hydrogen via elastic nuclear scattering in the conversion layer, and protons are produced as recoils. These protons enter the sensitive volume of the detector and are detected. In the process of detector development, Monte Carlo computer codes are necessary to simulate the detection process. This paper presents the main features of the S3M code (SRIM Supporting Software Modules) and shows its application potential. Examples are given for the neutron detectors with a conversion layer and for CVD (Chemical Vapor Deposition) diamond detectors for beam-condition monitors at the LHC (Large Hadron Collider). Special attention is paid to the S3M statistical modules that can be of interest also for other application areas like beam transport, accelerators, ion therapy, etc. The results are generated by MCNPX (Monte Carlo N-Particle eXtended) simulations used to optimize the thickness of the HDPE (high density polyethylene) conversion layer.

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Muscato, O., S. Rinaudo, and P. Falsaperla. "Calibration of a One Dimensional Hydrodynamic Simulator with Monte Carlo Data." VLSI Design 8, no. 1-4 (January 1, 1998): 515–20. http://dx.doi.org/10.1155/1998/47467.

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In this paper we use the code Exemplar for matching a hydrodynamic 1D, time-dependent simulator and the transport coefficients obtained by the Monte Carlo simulator Damocles. This code is based on the Least Square method and it does not require any a priori knowledge about the simulator (analytical form of the equations etc.). The stationary electron flow in a one dimensional n+−n−n+ submicron silicon diode is simulated.

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Sibczyński, Paweł, Andrzej Brosławski, Szymon Burakowski, Arkadiusz Chłopik, Marek Dryll, Andrzej Dziedzic, Michał Gierlik, et al. "Application of fluorine-based threshold activation detector for neutron flux calculation from D-T neutron generator." EPJ Web of Conferences 225 (2020): 02004. http://dx.doi.org/10.1051/epjconf/202022502004.

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In this paper we propose a method of fast neutron flux estimation from a pulsed D-T neutron generator with application of single CaF2 scintillation crystal. The analysis method relies on 19F(n, α)16N threshold activation reaction having neutron energy threshold at 1.6 MeV. As a result, the 16N undergo β− decay with half-life of 7.1 s, emitting β particles with endpoint up to 10.4 MeV in the scintillator medium. Integration of the β distribution curve, preceded by calculation of (n, α) rate on F with Monte Carlo N-Particle Transport Code v6 (MCNP6) for fixed geometry, allows to estimate the neutron flux in 4π per second within few minutes.

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Mancusi, Davide, Alice Bonin, François-Xavier Hugot, and Fadhel Malouch. "Advances in Monte-Carlo code TRIPOLI-4®’s treatment of the electromagnetic cascade." EPJ Web of Conferences 170 (2018): 01008. http://dx.doi.org/10.1051/epjconf/201817001008.

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TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

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Kim, Jae Seok, Kyoung Won Jang, Sang Hun Shin, Seon Guen Kim, Seung Han Hong, Hyeok In Sim, Jae Seok Jang, et al. "Estimation of Cerenkov Radiation Generated in Various Dielectrics Using MCNPX Simulation." Advanced Materials Research 1033-1034 (October 2014): 1127–30. http://dx.doi.org/10.4028/www.scientific.net/amr.1033-1034.1127.

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To estimate Cerenkov radiation generated in various dielectric materials, in this study, we calculated therapeutic photon beams induced electron fluxes in a CaF2, PMMA, SiO2 and Al2O3 by using the Monte Carlo N-particle Extended transport code (MCNPX). Also, we clarified the relationship between electron fluxes produced in various dielectrics and energy depositions in water by irradiation of therapeutic photon beams. The electron fluxes and the energy depositions were calculated as a function of water depth.

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HUGOT, Francois-Xavier, and Yi-Kang LEE. "A New Prototype Display Tool for the Monte Carlo Particle Transport Code TRIPOLI-4." Progress in Nuclear Science and Technology 2 (October 1, 2011): 851–54. http://dx.doi.org/10.15669/pnst.2.851.

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Sedlačková, Katarína, Andrea Šagátová, Bohumír Zat'ko, Vladimír Nečas, Michael Solar, and Carlos Granja. "MCNPX simulations of the silicon carbide semiconductor detector response to fast neutrons from D–T nuclear reaction." International Journal of Modern Physics: Conference Series 44 (January 2016): 1660226. http://dx.doi.org/10.1142/s201019451660226x.

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Silicon Carbide (SiC) has been long recognized as a suitable semiconductor material for use in nuclear radiation detectors of high-energy charged particles, gamma rays, X-rays and neutrons. The nuclear interactions occurring in the semiconductor are complex and can be quantified using a Monte Carlo-based computer code. In this work, the MCNPX (Monte Carlo N-Particle eXtended) code was employed to support detector design and analysis. MCNPX is widely used to simulate interaction of radiation with matter and supports the transport of 34 particle types including heavy ions in broad energy ranges. The code also supports complex 3D geometries and both nuclear data tables and physics models. In our model, monoenergetic neutrons from D–T nuclear reaction were assumed as a source of fast neutrons. Their energy varied between 16 and 18.2 MeV, according to the accelerating voltage of the deuterons participating in D–T reaction. First, the simulations were used to calculate the optimum thickness of the reactive film composed of High Density PolyEthylene (HDPE), which converts neutral particles to charged particles and thusly enhancing detection efficiency. The dependency of the optimal thickness of the HDPE layer on the energy of the incident neutrons has been shown for the inspected energy range. Further, from the energy deposited by secondary charged particles and recoiled ions, the detector response was modeled and the effect of the conversion layer on detector response was demonstrated. The results from the simulations were compared with experimental data obtained for a detector covered by a 600 and 1300 [Formula: see text]m thick conversion layer. Some limitations of the simulations using MCNPX code are also discussed.

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Wang, Yuan, Miao Zhang, Tong Song, and Zhenqi Chang. "Design and fabrication of 125I seeds for brachytherapy using capillary-based microfluidic technique." Nukleonika 66, no. 2 (June 1, 2021): 55–60. http://dx.doi.org/10.2478/nuka-2021-0007.

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Abstract A new kind of 125I seeds with a core-shell structure were synthesized by an easy assembling–disassembling coaxial capillaries microfluidic device. The dose distribution of a 125I brachytherapy source fabricated by arranging six 125I seeds collinearly within a cylindrical titanium capsule was simulated by modelling the source in a water phantom using Monte Carlo N-Particle Transport code. The influence of the motion and the core size of the 125I seeds on the dose distribution was also studied in this work.

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Asni, H., H. Wagiran, I. Hossain, A. T. Ramli, and M. I. Saripan. "Thermoluminescence energy response of TLD-100 subjected to photon irradiation using Monte Carlo N-particle transport code version 5." Journal of Engineering Thermophysics 20, no. 3 (August 3, 2011): 329–33. http://dx.doi.org/10.1134/s1810232811030118.

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Iwamoto, Hiroki, Alexey Stakovskiy, Luca Fiorito, and Gert Van den Eynde. "Sensitivity and uncertainty analysis of βeff for MYRRHA using a Monte Carlo technique." EPJ Nuclear Sciences & Technologies 4 (2018): 42. http://dx.doi.org/10.1051/epjn/2018023.

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This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction βeff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP. The βeff sensitivities are calculated by the modified k-ratio method proposed by Chiba. Comparing the βeff sensitivities obtained with different scaling factors a introduced by Chiba shows that a value of a = 20 is the most suitable for the uncertainty quantification of βeff. Using the calculated βeff sensitivities and the JENDL-4.0u covariance data, the βeff uncertainties for the critical and subcritical cores are determined to be 2.2 ± 0.2% and 2.0 ± 0.2%, respectively, which are dominated by delayed neutron yield of 239Pu and 238U.

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Lagaki, V., E. Kouvaris, and T. J. Mertzimekis. "A new γ-spectroscopy station at the University of Athens." HNPS Proceedings 22 (March 8, 2019): 126. http://dx.doi.org/10.12681/hnps.1919.

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A new γ-spectroscopy station at the University of Athens has been recently deployed. The station is built around a 25% HPGe detector and aims at supporting the environmental radioactivity studies program of the NuSTRAP group at UoA. The detector needs detailed characterization of its efficiency and calibration over a rather wide range of energies. Besides standard calibration point sources (60Co, 137Cs, 152Eu and 226Ra), detector simulations using the Monte Carlo N-particle transport code (MCNP) were performed. Results of the MCNP calculations are shown and compared with real spectra showing satisfactory agreement

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Stanković, Srboljub J., R. D. Ilić, O. Ciraj-Bjelac, M. Kovačević, and David Davidović. "Characterization of Target Material for X-Ray Generator by Monte Carlo Method." Materials Science Forum 555 (September 2007): 137–40. http://dx.doi.org/10.4028/www.scientific.net/msf.555.137.

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The adequate choice of different target materials for X-ray generators is a very important subject of engineers’ practice and research. In the present work we analyze theoretically the transport of electrons through the anode material and the production of the corresponding bremsstrahlung radiation. In our analysis we simulate the particle transport with the help of the FOTELP code, which is based on the Monte Carlo simulation. Our main aim is to develop an efficient and handy method, which could be helpful in improving the design of the X-ray tube components and in reducing of the patient dose, while keeping the image quality. The obtained results are encouraging.

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Bilalodin, Bilalodin, Gede Bayu Suparta, Arief Hermanto, Dwi Satya Palupi, and Yohannes Sardjono. "Characteristics in Water Phantom of Epithermal Neutron Beam Produced by Double Layer Beam Shaping Assembly." ASEAN Journal on Science and Technology for Development 36, no. 1 (April 27, 2019): 9–12. http://dx.doi.org/10.29037/ajstd.519.

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A Double Layer Beam Shaping Assembly (DLBSA) was designed to produce epithermal neutrons for BNCT purposes. The Monte Carlo N-Particle eXtended program was used as the software to design the DLBSA and phantom. Distribution of epithermal neutron and gamma flux in the DLBSA and phantom and absorbed dose in the phantom were computed using the Particle and Heavy Ion Transport code System program. Testing results of epithermal neutron beam irradiation of the water phantom showed that epithermal neutrons were thermalized and penetrated the phantom up to a depth of 12 cm. The maximum value of the absorbed dose was 2 × 10-3 Gy at a depth of 2 cm in the phantom.

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Mendoza, Alberto, Carlos Torres-Verdín, and Bill Preeg. "Linear iterative refinement method for the rapid simulation of borehole nuclear measurements: Part I — Vertical wells." GEOPHYSICS 75, no. 1 (January 2010): E9—E29. http://dx.doi.org/10.1190/1.3267877.

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As a result of its high numerical accuracy and versatility to include complex tool configurations and arbitrary spatial distributions of material properties, the Monte Carlo method is the foremost numerical technique used to simulate borehole nuclear measurements. Although recent advances in computer technology have considerably reduced the computer time required by Monte Carlo simulations of borehole nuclear measurements, the efficiency of the method is still not sufficient for estimation of layer-by-layer properties or combined quantitative interpretation with other borehole measurements. We develop and successfully test a new linear iterative refinement method to simulate nuclear borehole measurements accurately and rapidly. The approximation stems from Monte Carlo-derived geometric response factors, referred to as flux sensitivity functions (FSFs), for specific density and neutron-tool configurations. Our procedure first invokes the integral representation of Boltzmann’s transport equation to describe the detector response from the flux of particles emitted by the radioactive source. Subsequently, we use theMonte Carlo N-particle (MCNP) code to calculate the associated detector response function and the particle flux included in the integral form of Boltzmann’s equation. The linear iterative refinement method accounts for variations of the response functions attributable to local perturbations when numerically simulating neutron and density porosity logs. We quantify variations in the FSFs of neutron and density measurements from borehole environmental effects and spatial variations of formation properties. Simulations performed with the new approximations yield errors in the simulated value of density of less than [Formula: see text] with respect to Monte Carlo-simulated logs. Moreover, for the case of radial geometric factor of density, we observe a maximum shift of [Formula: see text] at 90% of the total sensitivity as a result of realistic variations of formation density. For radial variation of neutron properties (migration length), the maximum change in the radial length of investigation is [Formula: see text]. Neutron porosity values simulated with the new approximation differ by less than 10% from Monte Carlo simulations. The approximations enable the simulation of borehole nuclear measurements in seconds of CPU time compared to several hours with MCNP.

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Lüley, Jakub, Branislav Vrban, Štefan Čerba, Filip Osuský, and Vladimír Nečas. "Processing of the multigroup cross-sections for MCNP calculations." Nuclear Science and Technology 9, no. 2 (June 15, 2019): 17–24. http://dx.doi.org/10.53747/jnst.v9i2.48.

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Stochastic Monte Carlo (MC) neutron transport codes are widely used in various reactorphysics applications, traditionally related to criticality safety analyses, radiation shielding and validation of deterministic transport codes. The main advantage of Monte Carlo codes lies in their ability to model complex and detail geometries without the need of simplifications. Currently, one of the most accurate and developed stochastic MC code for particle transport simulation is MCNP. To achieve the best real world approximations, continuous-energy (CE) cross-section (XS) libraries are often used. These CE libraries consider the rapid changes of XS in the resonance energy range; however, computing-intensive simulations must be performed to utilize this feature. To broaden ourcomputation abilities for industrial application and partially to allow the comparison withdeterministic codes, the CE cross section library of the MCNP code is replaced by the multigroup (MG) cross-section data. This paper is devoted to the cross-section processing scheme involving modified versions of TRANSX and CRSRD codes. Following this approach, the same data may be used in deterministic and stochastic codes. Moreover, using formerly developed and upgraded crosssection processing scheme, new MG libraries may be tailored to the user specific applications. For demonstration of the proposed cross-section processing scheme, the VVER-440 benchmark devoted to fuel assembly and pip-by-pin power distribution was selected. The obtained results are compared with continues energy MCNP calculation and multigroup KENO-VI calculation.

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Chantzi, Stefania, Emmanouil Papanastasiou, Christina Athanasopoulou, Elisavet Molyvda-Athanasopoulou, Panagiotis Bamidis, and Anastasios Siountas. "Design of a Monte Carlo model based on dual-source computed tomography (DSCT) scanners for dose and image quality assessment using the Monte Carlo N-Particle (MCNP5) code." Polish Journal of Medical Physics and Engineering 26, no. 1 (March 1, 2020): 11–20. http://dx.doi.org/10.2478/pjmpe-2020-0002.

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AbstractThe purpose of this work was to develop and validate a Monte Carlo model for a Dual Source Computed Tomography (DSCT) scanner based on the Monte Carlo N-particle radiation transport computer code (MCNP5). The geometry of the Siemens Somatom Definition CT scanner was modeled, taking into consideration the x-ray spectrum, bowtie filter, collimator, and detector system. The accuracy of the simulation from the dosimetry point of view was tested by calculating the Computed Tomography Dose Index (CTDI) values. Furthermore, typical quality assurance phantoms were modeled in order to assess the imaging aspects of the simulation. Simulated projection data were processed, using the MATLAB software, in order to reconstruct slices, using a Filtered Back Projection algorithm. CTDI, image noise, CT-number linearity, spatial and low contrast resolution were calculated using the simulated test phantoms. The results were compared using several published values including IMPACT, NIST and actual measurements. Bowtie filter shapes are in agreement with those theoretically expected. Results show that low contrast and spatial resolution are comparable with expected ones, taking into consideration the relatively limited number of events used for the simulation. The differences between simulated and nominal CT-number values were small. The present attempt to simulate a DSCT scanner could provide a powerful tool for dose assessment and support the training of clinical scientists in the imaging performance characteristics of Computed Tomography scanners.

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Griesheimer, D. P., D. F. Gill, B. R. Nease, T. M. Sutton, M. H. Stedry, P. S. Dobreff, D. C. Carpenter, et al. "MC21 v.6.0 – A continuous-energy Monte Carlo particle transport code with integrated reactor feedback capabilities." Annals of Nuclear Energy 82 (August 2015): 29–40. http://dx.doi.org/10.1016/j.anucene.2014.08.020.

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He, Qingming, Qi Zheng, Jie Li, Hongchun Wu, Wei Shen, Liangzhi Cao, Zhouyu Liu, and Jialong Xu. "NECP-MCX: A hybrid Monte-Carlo-Deterministic particle-transport code for the simulation of deep-penetration problems." Annals of Nuclear Energy 151 (February 2021): 107978. http://dx.doi.org/10.1016/j.anucene.2020.107978.

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Kalospyros, Spyridon A., Violeta Gika, Zacharenia Nikitaki, Antigoni Kalamara, Ioanna Kyriakou, Dimitris Emfietzoglou, Michael Kokkoris, and Alexandros G. Georgakilas. "Monte Carlo Simulation-Based Calculations of Complex DNA Damage for Incidents of Environmental Ionizing Radiation Exposure." Applied Sciences 11, no. 19 (September 27, 2021): 8985. http://dx.doi.org/10.3390/app11198985.

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In this paper, we present a useful Monte Carlo (MC)-based methodology that can be utilized to calculate the absorbed dose and the initial levels of complex DNA damage (such as double strand breaks-DSBs) in the case of an environmental ionizing radiation (IR) exposure incident (REI) i.e., a nuclear accident. Our objective is to assess the doses and complex DNA damage by isolating only one component of the total radiation released in the environment after a REI that will affect the health of the exposed individual. More specifically, the radiation emitted by radionuclide 137Cs in the ground (under the individual’s feet). We use a merging of the Monte Carlo N-Particle Transport code (MCNP) with the Monte Carlo Damage Simulation (MCDS) code. The DNA lesions have been estimated through simulations for different surface activities of a 137Cs ground-based γ radiation source. The energy spectrum of the emitted secondary electrons and the absorbed dose in typical mammalian cells have been calculated using the MCNP code, and then these data are used as an input in the MCDS code for the estimation of critical DNA damage levels and types. As a realistic application, the calculated dose is also used to assess the Excess Lifetime Cancer Risk (ELCR) for eight hypothetical individuals, living in different zones around the Chernobyl Nuclear Power Plant, exposed to different time periods at the days of the accident in 1986. We conclude that any exposition of an individual in the near zone of Chernobyl increases the risk of cancer at a moderate to high grade, connected also with the induction of complex DNA damage by radiation. Generally, our methodology has proven to be useful for assessing γ rays-induced complex DNA damage levels of the exposed population, in the case of a REI and for better understanding the long-term health effects of exposure of the population to IR.

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Li, Jun Heng, Rong Hua Huang, and Hao Ran Cao. "Static Neutronics Analyses of Hellium-Cooled Solid Breeder Blanket." Advanced Materials Research 953-954 (June 2014): 631–34. http://dx.doi.org/10.4028/www.scientific.net/amr.953-954.631.

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A Monte Carlo N-particle transport code was used to study physics of the helium-cooled solid-breeder tritium breeding blanket in the Chinese Fusion Engineering Thermal Reactor （CFETR）for various volume ratio of the neutron multiplier and tritium breeder and various thickness of the first wall. A sandwich-type of Be and loading model is used to analyze the compact of volume ratio and the thickness of the first wall for the tritium breeding rate. The results of different volume ratio models show that the tritium breeding ratio would reach 1.51 for volume ratio from 2 to 5.And the results of the different first wall thickness show that the upper limit of the thickness should be 33mm to keep the tritium self-sustain.

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Toker, Ozan, Bayram Bilmez, H. Birtan Kavanoz, Özgür Akçalı, and Orhan İçelli. "Comparison of ITO and ZnO ternary glassy composites in terms of radiation shielding properties by Monte Carlo N-particle transport code and BXCOM." Radiation and Environmental Biophysics 59, no. 2 (March 19, 2020): 283–93. http://dx.doi.org/10.1007/s00411-020-00838-x.

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Silva, C. A. M., J. A. D. Salomé, B. T. Guerra, C. Pereira, A. L. Costa, M. A. F. Veloso, M. A. B. C. Menezes, and H. M. Dalle. "Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code." International Journal of Nuclear Energy 2014 (September 9, 2014): 1–9. http://dx.doi.org/10.1155/2014/793934.

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In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport) code. The sensitivity analyses included small differences of the core and the rods dimensions and different levels of model detailing. Four models were simulated and neutronic parameters such as effective multiplication factor (keff), reactivity (ρ), and thermal and total neutron flux in central thimble in some different conditions of the reactor operation were analysed. The simulated models presented good agreement between them, as well as in comparison with available experimental data. In this way, the sensitivity analyses demonstrated that simulations of the TRIGA IPR-R1 reactor can be performed using any one of the four investigated MCNP models to obtain the referenced neutronic parameters.

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Yegin, Gultekin. "A new approach to geometry modeling for Monte Carlo particle transport: An application to the EGS code system." Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms 211, no. 3 (November 2003): 331–38. http://dx.doi.org/10.1016/s0168-583x(03)01318-1.

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Zaman, Fahad A., Lawrence W. Townsend, Wouter C. de Wet, and Naser T. Burahmah. "The Lunar Radiation Environment: Comparisons between PHITS, HETC-HEDS, and the CRaTER Instrument." Aerospace 8, no. 7 (July 8, 2021): 182. http://dx.doi.org/10.3390/aerospace8070182.

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Understanding the radiation environment near the lunar surface is a key step towards planning for future missions to the Moon. However, the complex variety of energies and particle types constituting the space radiation environment makes the process of replicating such environment very difficult in Earth-based laboratories. Radiation transport codes provide a practical alternative covering a wider range of particle energy, angle, and type than can be experimentally attainable. Comparing actual measurements with simulation results help in validating particle flux input models, and input collision models and databases involving nuclear and electromagnetic interactions. Thus, in this work, we compare the LET spectra simulated using the Monte Carlo transport code PHITS with measurements made by the CRaTER instrument that is currently orbiting the Moon studying its radiation environment. In addition, we utilize a feature in PHITS that allows the user to run the simulations without Vavilov energy straggling to test whether it is the root cause of erroneous phenomena exhibited in similar studies in literature. The results herein show good agreement between the LET spectra of PHITS and the CRaTER instrument. They also confirm that using a Vavilov distribution correction would ultimately provide a better agreement between CRaTER measurements and the previous LET spectra from the transport codes HETC-HEDS and HZETRN.

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Abdul Haneefa, K., T. Siji Cyriac, M. M. Musthafa, R. Ganapathi Raman, V. T. Hridya, A. Siddhartha, and K. K. Shakir. "FLUKA Monte Carlo for Basic Dosimetric Studies of Dual Energy Medical Linear Accelerator." Journal of Radiotherapy 2014 (July 24, 2014): 1–7. http://dx.doi.org/10.1155/2014/343979.

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General purpose Monte Carlo code for simulation of particle transport is used to study the basic dosimetric parameters like percentage depth dose and dose profiles and compared with the experimental measurements from commercial dual energy medical linear accelerator. Varian Clinac iX medical linear accelerator with dual energy photon beams (6 and 15 MV) is simulated using FLUKA. FLAIR is used to visualize and edit the geometry. Experimental measurements are taken for 100 cm source-to-surface (SSD) in 50 × 50 × 50 cm3 PTW water phantom using 0.12 cc cylindrical ionization chamber. Percentage depth dose for standard square field sizes and dose profiles for various depths are studied in detail. The analysis was carried out using ROOT (a DATA analysis frame work developed at CERN) system. Simulation result shows good agreement in percentage depth dose and beam profiles with the experimental measurements for Varian Clinac iX dual energy medical linear accelerator.

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Saraniti, Marco, Yibing Hu, and Stephen M. Goodnick. "Particle-based Full-band Approach for Fast Simulation of Charge Transport in Si, GaAs, and InP." VLSI Design 15, no. 4 (January 1, 2002): 743–50. http://dx.doi.org/10.1080/1065514021000012354.

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We discuss the application of the fullband cellular automaton (CA) method for the simulation of charge transport in several semiconductors. Basing the selection of the state after scattering on simple look-up tables, the approach is physically equivalent to the full band Monte Carlo (MC) approach but is much faster. Furthermore, the structure of the pre-tabulated transition probabilities naturally allows for an extension of the model to fully anisotropic scattering without additional computational burden. Simulation results of transport of electrons and holes in several materials are discussed, with particular emphasis on the transient response of photo-generated carriers in InP and GaAs. Finally, a discussion on parallel algorithms is presented, for the implementation of the code on workstation clusters.

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Zheng, Qi, Wei Shen, Xuesong Li, Tengfei Hao, Qingming He, Jie Li, and Zhouyu Liu. "A HYBRID MONTE-CARLO-DETERMINISTIC METHOD FOR AP1000 EX-CORE DETECTOR RESPONSE SIMULATION." EPJ Web of Conferences 247 (2021): 05003. http://dx.doi.org/10.1051/epjconf/202124705003.

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The ex-core detector-response calculation is a typical deep-penetration problem, which is challenging for the Monte Carlo method. The response of the ex-core detector is an important parameter for the safe operation of the nuclear power plants. Meanwhile, evaluation of the ex-core detector response during each step of fuel-loading is used to guide the fuel-loading sequence. The response can also be used to reconstruct core-power distribution for online monitoring of long-term power. The detector used for the ex-core response is the source-range detector which is sensitive to thermal neutrons. For a Monte Carlo shielding calculation of the above detector response, the thermal flux under 0.625eV is needed, which is too small to be tallied by traditional Monte Carlo simulations. In practice, the tally results are close to zero in the detector region under direct Monte Carlo calculation. Even if the number of particles is increased to a significant amount, the statistical variance is still very large. The high variance along with a significant calculation time leads to a small Figure Of Merit (FOM). In order to solve this problem and to improve the tally efficiency of the ex-core detector response, a hybrid Monte-Carlo-deterministic method is employed in this study, and an in-house hybrid Monte-Carlo-deterministic particle transport code, NECP-MCX, is developed in this paper. The method takes the space-energy-dependent adjoint fluxes to generate importance parameters for the mesh-based weight window in the Monte Carlo calculation. Simultaneously, the mesh-based source biasing is performed with the consistent importance parameters to make the starting weight of neutrons matching with the survival weight of the weight windows. As the mesh used in the hybrid Monte-Carlo-deterministic method is superimposed, the mesh of the weight window will not be affected by the complex geometry model. The adjoint flux is obtained by the efficient SN method with the multi-group cross-section data. The whole toolset is convenient to use with single set of the modelling data for both Monte Carlo and deterministic simulations. Compared with the direct Monte Carlo simulation, the hybrid Monte-Carlo-deterministic method has a higher efficiency for a typical deep-penetration problem such as the AP1000 ex-core detector-response simulation.

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Iwamoto, Yosuke, and Tatsuhiko Ogawa. "Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation." Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms 396 (April 2017): 26–33. http://dx.doi.org/10.1016/j.nimb.2017.02.007.

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Batmunkh, Munkhbaatar, Lkhagvaa Bayarchimeg, Aleksandr N. Bugay, and Oidov Lkhagva. "Monte Carlo track structure simulation in studies of biological effects induced by accelerated charged particles in the central nervous system." EPJ Web of Conferences 204 (2019): 04008. http://dx.doi.org/10.1051/epjconf/201920404008.

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Simulating the biological damage induced by charged particles trajectories (tracks) in the central nervous system (CNS) at different levels of its organization (molecular, cellular, and tissue) is a challenge of modern radiobiology studies. According to the recent experimental studies at particle accelerators, the most radiation-sensitive area of the CNS is the hippocampus. In this regards, the development of measurement-based Monte Carlo simulation of radiation-induced alterations in the hippocampus is of great interest to understand the radiobiological effects on the CNS. The present work investigates the influence of charged particles on the hippocampal cells of the rat brain using the Geant4 Monte Carlo radiation transport code. The applied computer simulation provides a method to simulate physics processes and chemical reactions in the developed model of the rat hippocampus, which contains different types of neural cells - pyramidal cells, mature and immature granular cells, mossy cells, and neural stem cells. The distribution of stochastic energy depositions has been obtained and analyzed in critical structures of the hippocampal neurons after irradiation with 600 MeV/u iron particles. The computed energy deposition in irradiated hippocampal neurons following a track of iron ion suggests that most of the energy is accumulated by granular cells. The obtained quantities at the level of molecular targets also assume that NMDA and GABA receptors belong to the most probable targets in the irradiated neural cells.

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Journal articles on the topic 'Monte Carlo N-Particle Transport Code'

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