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Title: Monte Carlo calculations of thermal neutron capture in gadolinium: A comparison of GEANT4 and MCNP with measurements

Abstract

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S({alpha},{beta})] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S({alpha},{beta}). The location of the thermal neutron peak calculated with MCNP without S({alpha},{beta}) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not includedmore » in GEANT4 it is not suitable for NCT applications.« less

Authors:
; ; ;  [1];  [2];  [3];  [2]
  1. Division of Biomedical Radiation Sciences, Department of Oncology, Radiology and Clinical Immunology, Rudbeck Laboratory, Uppsala University, SE-751 85 Uppsala (Sweden)
  2. (Sweden)
  3. (Sweden) and Studsvik Medical AB, SE-612 82 Nykoeping (Sweden)
Publication Date:
OSTI Identifier:
20775057
Resource Type:
Journal Article
Resource Relation:
Journal Name: Medical Physics; Journal Volume: 33; Journal Issue: 2; Other Information: DOI: 10.1118/1.2150787; (c) 2006 American Association of Physicists in Medicine; Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
62 RADIOLOGY AND NUCLEAR MEDICINE; A CODES; BIOLOGICAL MATERIALS; GADOLINIUM; MONTE CARLO METHOD; NEUTRON CAPTURE THERAPY; PHANTOMS; RADIATION DOSE DISTRIBUTIONS; THERMAL NEUTRONS

Citation Formats

Enger, Shirin A., Munck af Rosenschoeld, Per, Rezaei, Arash, Lundqvist, Hans, Department of Radiation Physics, Lund University Hospital, SE-22185 Lund, Division of Medical Radiation Physics, Department of Oncology-Pathology, Karolinska Institutet, SE-171 76 Stockholm, and Division of Biomedical Radiation Sciences, Department of Oncology, Radiology and Clinical Immunology, Rudbeck Laboratory, Uppsala University, SE-751 85 Uppsala. Monte Carlo calculations of thermal neutron capture in gadolinium: A comparison of GEANT4 and MCNP with measurements. United States: N. p., 2006. Web. doi:10.1118/1.2150787.
Enger, Shirin A., Munck af Rosenschoeld, Per, Rezaei, Arash, Lundqvist, Hans, Department of Radiation Physics, Lund University Hospital, SE-22185 Lund, Division of Medical Radiation Physics, Department of Oncology-Pathology, Karolinska Institutet, SE-171 76 Stockholm, & Division of Biomedical Radiation Sciences, Department of Oncology, Radiology and Clinical Immunology, Rudbeck Laboratory, Uppsala University, SE-751 85 Uppsala. Monte Carlo calculations of thermal neutron capture in gadolinium: A comparison of GEANT4 and MCNP with measurements. United States. doi:10.1118/1.2150787.
Enger, Shirin A., Munck af Rosenschoeld, Per, Rezaei, Arash, Lundqvist, Hans, Department of Radiation Physics, Lund University Hospital, SE-22185 Lund, Division of Medical Radiation Physics, Department of Oncology-Pathology, Karolinska Institutet, SE-171 76 Stockholm, and Division of Biomedical Radiation Sciences, Department of Oncology, Radiology and Clinical Immunology, Rudbeck Laboratory, Uppsala University, SE-751 85 Uppsala. Wed . "Monte Carlo calculations of thermal neutron capture in gadolinium: A comparison of GEANT4 and MCNP with measurements". United States. doi:10.1118/1.2150787.
@article{osti_20775057,
title = {Monte Carlo calculations of thermal neutron capture in gadolinium: A comparison of GEANT4 and MCNP with measurements},
author = {Enger, Shirin A. and Munck af Rosenschoeld, Per and Rezaei, Arash and Lundqvist, Hans and Department of Radiation Physics, Lund University Hospital, SE-22185 Lund and Division of Medical Radiation Physics, Department of Oncology-Pathology, Karolinska Institutet, SE-171 76 Stockholm and Division of Biomedical Radiation Sciences, Department of Oncology, Radiology and Clinical Immunology, Rudbeck Laboratory, Uppsala University, SE-751 85 Uppsala},
abstractNote = {GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S({alpha},{beta})] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S({alpha},{beta}). The location of the thermal neutron peak calculated with MCNP without S({alpha},{beta}) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.},
doi = {10.1118/1.2150787},
journal = {Medical Physics},
number = 2,
volume = 33,
place = {United States},
year = {Wed Feb 15 00:00:00 EST 2006},
month = {Wed Feb 15 00:00:00 EST 2006}
}
  • First-principles NaI and BGO detector response functions calculations made with the MCNP code are compared to measurements. Excellent agreement is achieved for the experiments analyzed. Such calculational methodology can be used to achieve a better understanding of the physics of detector response and to maximize the information content available from measured data.
  • A comparison is reported of measured and Monte Carlo calculated spatial distributions of U/sup 238/ resonance neutron capture in a 0.387-in. diameter uranium metal rod in a hexagonal light-water moderated lattice with rod centerto- center spacing of 0.567 in. The following calculational results were obtained: (1) the fraction of neutrons started at 10 kev which are captured in the rod above 3 ev; (2) the fraction of total capture in each of the twenty rings across the rod; and (3) the fraction of total capture in each of 16 energy intervals for each of the two main regions of themore » rod. Curves are given which show the comparison of the adjusted Monte Carlo calculation with the measured spatial distribution. These results indicate that the code used is quite capable of giving a detailed prediction of the spatial distribution of resonance neutron capture across the rod diameter. It is significant to note that the agreement between experiment and calculation gives added justification to the treatment of the higher energy resonances. (B.O.G.)« less
  • Purpose: Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on themore » correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. Methods: For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Results: Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons.Conclusion: Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.« less