CHARACTERIZATION OF THE GAMMA DOSE COMPONENT IN THE NEUTRON FIELD OF A BNCT IRRADIATION FACILITY

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Abstract

At the University of Pavia, a neutron irradiation facility has been constructed for preclinical in vitro and in vivo studies for BNCT. The facility is a wide channel (40 x 20 x 100 cm3) inside the graphite Thermal Column of the 250 KW research reactor TRIGA Mark II. The neutron field consists in a thermal component ranging from about 1010 and 109 cm-2s-1 along the longitudinal axis of the channel and it is uniform within 10% along the transversal axes. The fast neutron contamination (En > 1.58 keV) is more than 2 orders of magnitude lower. The gamma background coming from the core has been reduced with a 20 cm thick bismuth shield, however in the facility a gamma component is still present due to neutron capture reactions in Bi and C. A set of simulations with the Monte Carlo code MCNP has been conducted together with measurements using alanine dosimeters to determine the gamma background during irradiation. Attention has been put in the choice of the material surrounding the alanine dosimeters (r = 2.4 mm and height h = 30 mm) in order to ensure electronic equilibrium: hydrogenated materials have been discarded to avoid additional dose due to the (n, γ) reactions in H: graphite holders have been used to this end. Simulations have been performed to choose the calibration gamma source: the gamma spectrum present in the facility has been calculated as well as the spectra of energy deposition in alanine due to the charged secondary radiation (electrons). A photon source produced in a 6 MV electrons linear accelerator was found to be much more suitable than the 60Co source typically employed to calibrate dosimeters. To determine the thermal neutron contribution to the alanine response a lithium carbonate (Li-6 enrichment: 95%) shield (attenuation factor of about three orders of magnitude) was prepared to host the dosimeters during irradiation; measurements with and without shield were done; gamma dose rates were between 6.0 ± 0.3 Gy/h and 1.5 ± 0.1 Gy/h along the longitudinal axis of the Column. The comparison between the experimental results and the simulations allowed determining a sensitivity factor of alanine to thermal and fast neutron dose equal to 0.4.
Original languageEnglish
Number of pages1
Publication statusPublished - 2017

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alanine
dosimeters
neutrons
dosage
irradiation
fast neutrons
thermal neutrons
graphite
electron radiation
simulation
neutron irradiation
linear accelerators
holders
bismuth
carbonates
contamination
lithium
attenuation
reactors
sensitivity

Cite this

@conference{2d3bb225b3be4ca88ee99d8e336ddc57,
title = "CHARACTERIZATION OF THE GAMMA DOSE COMPONENT IN THE NEUTRON FIELD OF A BNCT IRRADIATION FACILITY",
abstract = "At the University of Pavia, a neutron irradiation facility has been constructed for preclinical in vitro and in vivo studies for BNCT. The facility is a wide channel (40 x 20 x 100 cm3) inside the graphite Thermal Column of the 250 KW research reactor TRIGA Mark II. The neutron field consists in a thermal component ranging from about 1010 and 109 cm-2s-1 along the longitudinal axis of the channel and it is uniform within 10{\%} along the transversal axes. The fast neutron contamination (En > 1.58 keV) is more than 2 orders of magnitude lower. The gamma background coming from the core has been reduced with a 20 cm thick bismuth shield, however in the facility a gamma component is still present due to neutron capture reactions in Bi and C. A set of simulations with the Monte Carlo code MCNP has been conducted together with measurements using alanine dosimeters to determine the gamma background during irradiation. Attention has been put in the choice of the material surrounding the alanine dosimeters (r = 2.4 mm and height h = 30 mm) in order to ensure electronic equilibrium: hydrogenated materials have been discarded to avoid additional dose due to the (n, γ) reactions in H: graphite holders have been used to this end. Simulations have been performed to choose the calibration gamma source: the gamma spectrum present in the facility has been calculated as well as the spectra of energy deposition in alanine due to the charged secondary radiation (electrons). A photon source produced in a 6 MV electrons linear accelerator was found to be much more suitable than the 60Co source typically employed to calibrate dosimeters. To determine the thermal neutron contribution to the alanine response a lithium carbonate (Li-6 enrichment: 95{\%}) shield (attenuation factor of about three orders of magnitude) was prepared to host the dosimeters during irradiation; measurements with and without shield were done; gamma dose rates were between 6.0 ± 0.3 Gy/h and 1.5 ± 0.1 Gy/h along the longitudinal axis of the Column. The comparison between the experimental results and the simulations allowed determining a sensitivity factor of alanine to thermal and fast neutron dose equal to 0.4.",
author = "Salvatore Gallo and Maurizio Marrale",
year = "2017",
language = "English",

}

TY - CONF

T1 - CHARACTERIZATION OF THE GAMMA DOSE COMPONENT IN THE NEUTRON FIELD OF A BNCT IRRADIATION FACILITY

AU - Gallo, Salvatore

AU - Marrale, Maurizio

PY - 2017

Y1 - 2017

N2 - At the University of Pavia, a neutron irradiation facility has been constructed for preclinical in vitro and in vivo studies for BNCT. The facility is a wide channel (40 x 20 x 100 cm3) inside the graphite Thermal Column of the 250 KW research reactor TRIGA Mark II. The neutron field consists in a thermal component ranging from about 1010 and 109 cm-2s-1 along the longitudinal axis of the channel and it is uniform within 10% along the transversal axes. The fast neutron contamination (En > 1.58 keV) is more than 2 orders of magnitude lower. The gamma background coming from the core has been reduced with a 20 cm thick bismuth shield, however in the facility a gamma component is still present due to neutron capture reactions in Bi and C. A set of simulations with the Monte Carlo code MCNP has been conducted together with measurements using alanine dosimeters to determine the gamma background during irradiation. Attention has been put in the choice of the material surrounding the alanine dosimeters (r = 2.4 mm and height h = 30 mm) in order to ensure electronic equilibrium: hydrogenated materials have been discarded to avoid additional dose due to the (n, γ) reactions in H: graphite holders have been used to this end. Simulations have been performed to choose the calibration gamma source: the gamma spectrum present in the facility has been calculated as well as the spectra of energy deposition in alanine due to the charged secondary radiation (electrons). A photon source produced in a 6 MV electrons linear accelerator was found to be much more suitable than the 60Co source typically employed to calibrate dosimeters. To determine the thermal neutron contribution to the alanine response a lithium carbonate (Li-6 enrichment: 95%) shield (attenuation factor of about three orders of magnitude) was prepared to host the dosimeters during irradiation; measurements with and without shield were done; gamma dose rates were between 6.0 ± 0.3 Gy/h and 1.5 ± 0.1 Gy/h along the longitudinal axis of the Column. The comparison between the experimental results and the simulations allowed determining a sensitivity factor of alanine to thermal and fast neutron dose equal to 0.4.

AB - At the University of Pavia, a neutron irradiation facility has been constructed for preclinical in vitro and in vivo studies for BNCT. The facility is a wide channel (40 x 20 x 100 cm3) inside the graphite Thermal Column of the 250 KW research reactor TRIGA Mark II. The neutron field consists in a thermal component ranging from about 1010 and 109 cm-2s-1 along the longitudinal axis of the channel and it is uniform within 10% along the transversal axes. The fast neutron contamination (En > 1.58 keV) is more than 2 orders of magnitude lower. The gamma background coming from the core has been reduced with a 20 cm thick bismuth shield, however in the facility a gamma component is still present due to neutron capture reactions in Bi and C. A set of simulations with the Monte Carlo code MCNP has been conducted together with measurements using alanine dosimeters to determine the gamma background during irradiation. Attention has been put in the choice of the material surrounding the alanine dosimeters (r = 2.4 mm and height h = 30 mm) in order to ensure electronic equilibrium: hydrogenated materials have been discarded to avoid additional dose due to the (n, γ) reactions in H: graphite holders have been used to this end. Simulations have been performed to choose the calibration gamma source: the gamma spectrum present in the facility has been calculated as well as the spectra of energy deposition in alanine due to the charged secondary radiation (electrons). A photon source produced in a 6 MV electrons linear accelerator was found to be much more suitable than the 60Co source typically employed to calibrate dosimeters. To determine the thermal neutron contribution to the alanine response a lithium carbonate (Li-6 enrichment: 95%) shield (attenuation factor of about three orders of magnitude) was prepared to host the dosimeters during irradiation; measurements with and without shield were done; gamma dose rates were between 6.0 ± 0.3 Gy/h and 1.5 ± 0.1 Gy/h along the longitudinal axis of the Column. The comparison between the experimental results and the simulations allowed determining a sensitivity factor of alanine to thermal and fast neutron dose equal to 0.4.

UR - http://hdl.handle.net/10447/241133

UR - https://neudos2017.ifj.edu.pl/index.html

M3 - Other

ER -