PubMed · 16585717
The source did not provide an abstract. Follow the original record for more information.
Explore connections, maps & timelines
A J Munro. 2006. Particle matters.. https://doi.org/10.1259/bjr%2F14849445
Cite the original work for its findings. Save a collection to share your selection of sources.
KEEP EXPLORING
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.
Two batches of diodes, with different structural ratios (the ratio of area and thickness), were made using different manufacturing processes. The energy response of the first batch to 15 kinds of monoenergetic neutrons ranging from 180 keV to 17.56 MeV was tested, and the neutron source response of both batches to 239Pu-Be neutron source was measured. The energy deposition in the diodes irradiated by 1 keV to 20 MeV monoenergetic neutrons was calculated with simulation procedure. The response curve of the experimental results showed an approximately similar trend to that of theoretical computation. Based on the results of the neutron source response experiments, it was concluded that the response of fast neutron varied linearly with the structural ratio of the detectors.
This paper describes the results of a study performed on a mixed field neutron/gamma (n/gamma) area dosemeter incorporating radiophotoluminescent (RPL) glass detectors. RPL glass is known to be virtually insensitive to neutrons. The aim of the study was therefore to determine the neutron response of a dosemeter designed to combine n/gamma conversion with RPL detection capability. Monte Carlo calculations as well as measurements using monoenergetic beams and isotopic neutron sources showed this response to be constant, to within 30% in terms of H*(10), and independent of neutron energy from 250 keV to 10 MeV. For area monitoring, tests carried out in nuclear facilities (around PuO2 glove box and shipping casks containing PWR, MOX spent fuels or vitrified fission product) demonstrated that dosemeter response was accurate to within 15%, where the gamma component of the mixed n,gamma field remained below 1 MeV. When exposed in the Silene reactor simulating a criticality accident (10(17) fissions-liquid 235U--e.g. 1 Gy neutron and 1 Gy photon), the dosemeter exhibited good correlation with reference values and other measurement technologies (again to within 30%), for both neutron and gamma absorbed dose.