Monitor de área pasivo para neutrones

La relevancia de la presencia de neutrones en los LINACs para radioterapia ha sido establecida en los reportes 79 y 102 de la NCRP1. La medición de la contribución de la dosis y su distribución en el paciente es importante para evaluar el riesgo de la inducción de tumores secundarios2. Para medir el espectro de neutrones o cualquier magnitud dosimétrica asociada, dentro de las salas de radioterapia con LINACs se debe usar un sistema pasivo3. En este trabajo se presenta el diseño de un monitor de área para neutrones con detector pasivo, así como su desempeño en un LINAC de 15 MV

Mini Subcritical Nuclear Reactor

A mini subcritical nuclear reactor was designed using Monte Carlo methods. The reactor has light water as moderator, natural uranium as fuel, and a 239PuBe neutron source. In the design uranium fuel was modeled in an arrangement of concentric rings: 8.5, 14.5, 20.5 26.5, 32.5 cm-inner radius, 3 cm-thick, and 36 cm-high. Different models were made from a single ring of natural uranium to five rings. For each case, the neutron spectra, the neutron fluence distribution, the effective multiplication factor, the amplification factor, and the reactor power were estimated. The ambient dose equivalent rate outside the mini reactor was also estimated. The maximum value for the keff (0.78) was obtained when five rings of fuel were used; this value is close to 0.86 which belongs to a Nuclear Chicago subcritical reactor which requires almost twice the amount of uranium than the mini subcritical reactor

Response to Neutrons and γ-rays of Two Liquid Scintillators

UltimaGoldTM AB and OptiphaseTrisafe are two liquid scintillators made by Perkin Elmer and EG & G Company respectively. Both are commercially promoted as scintillation detectors for α and β particles. In this work, the responses to γ-rays and neutrons of UltimaGoldTM AB and OptiphaseTriSafe liquid scintillators, without and with reflector, have been measured aiming to use these scintillators as γ-rays and neutron detectors. Responses to γ-rays and neutrons were measured as pulse shape spectra in a multichannel analyzer. Scintillators were exposed to gamma rays produced by 137Cs, 54Mn, 22Na and 60Co sources. The response to neutrons was obtained with a 241AmBe neutron source that was measured to 25 and 50 cm from the scintillators. The pulse height spectra due to gamma rays are shifted to larger channels as the photon energy increases and these responses are different from the response due to neutrons. Thus, UltimaGoldTM AB and OptiphaseTrisafe can be used to detect γ-rays and neutrons

Studies on neutron and photon kerma parameters for human body organs

A study on neutron kerma factors and photon air-kerma for human organs is presented for neutron energy range 2.53×10-8 MeV to 29 MeV and photon energy range 1 keV to 20 MeV. The human organs water equivalence for photon and neutron, is also presented. The ratio of the mass-energy absorption coefficients of human organs to water was found constant and unity above 100 keV, whereas there was a large difference for energies below 100 keV. The neutron kerma factors of human organs and water are found of same order of magnitude whereas differs for air. The neutron kerma factors of human organs and tissue substitutes were found to be equal to water for neutron energies between 63 eV and 200 keV. The skeleton-cortical bone was found to be away from water equivalence for low-energy photons and high-energy neutrons

Monte Carlo calculation of the response matrix of a Bonner spheres spectrometer

Abstract. The Bonner spheres spectrometer is utilized to estimate the neutron spectrum of neutrons from thermal up to several MeV neutrons. Its response is increased to few GeV neutrons by introducing large Z materials as inner shells. To use the spectrometer a matrix response and an unfolding method are required; these are crucial to assure the quality of spectrometer output. The response matrix of a Bonner sphere spectrometer was calculated by use of the MCNP code. As thermal neutron counter the spectrometer has a 0.4 Ø × 0.4 cm2 6 LiI(Eu) scintillator which is located at the centre of a set of polyethylene spheres. The response functions were calculated for 0, 2, 3, 5, 8, 10, and 12 inches-diameter polyethylene spheres for neutrons whose energy goes from 10 -8 to 100 MeV. For energies from 10 -8 to 20 MeV the MCNP4C code was utilized while for neutrons from 20 to 100 MeV calculations were carried out with MCNPX code. The response functions were compared with those reported in the literature

Feasibility of 18-MV grid therapy from radiation protection aspects: unwanted dose and fatal cancer risk caused by photoneutrons and scattered photons

Purpose: Photoneutron production is a common concern when using 18-MV photon beams in radiation therapy. In Spatially Fractionated Grid Radiation Therapy (SFGRT), the grid block in the collimation system modifies the neutron production, photon scattering, and electron contamination in and out of the radiation field. Such an effect was studied with grids made of different high-Z materials by Monte Carlo simulations. The results were also used to evaluate the lifetime risk of fatal cancers. Methods: MCNPX® code (2.7.0 extensions) was employed to simulate an 18-MV LINAC (Varian 2100 C/D). Three types of grid made of brass, cerrobend, and lead were used to study the neutron and electron fluence. Output factors for each grid with different field sizes were calculated. A revised female MIRD phantom with an 8-cm spherical tumor inside the liver was used to estimate the dose to the tumor and the critical organs. A 20-Gy SFGRT plan with Anterior Posterior (AP) - Posterior Anterior (PA) grid beams was compared with a Conventional Fractionated Radiation Therapy (CFRT) plan which delivered 40-Gy to the tumor by AP-PA open beams. Neutron equivalent dose, photon equivalent dose, as well as lifetime risks of fatal cancer were calculated in the organs at risk. Results: The grid blocks reduced the fluence of contaminant electrons inside the treatment field by more than 50%. The neutron fluences per electron-history in SFGRT plans with brass, cerrobend and lead were on average 55%, 31% and 31% less than that of the CFRT plan, respectively. However, when converting to fluences per delivered dose (Gy), the cerrobend and lead grid may incur higher neutron dose for 20 × 20 cm2 field size and above. The changes in neutron mean energy, as well as the correlated radiation weighting factors, were insignificant. The total risk due to the photoneutrons in the SFGRT plans was 87% or lower than that in the CFRT plans. In both SFGRT and CFRT plans, the contribution of the primary and scattered photons to the fatal cancer risk was 2 times or more than the photoneutrons. The total risks from photons in SFGRT with brass, cerrobend, and lead blocks were 1.733, 1.374, and 1.260%, respectively, which were less than 30% of the total photon-risk in CFRT (5.827%). Conclusion: In the brass, cerrobend, and lead grids, the attenuation of photoneutrons outweighs its photoneutron production in 18-MV SFGRT. The total cancer risks from photons and photoneutrons in the SFGRT plans were 30% or less of the risks in the CFRT plans (5.911%). Using 18 MV photon beams with brass, cerrobend, and lead grid blocks is still a feasible option for SFGRT

Dosimetric evaluation of 123I (Iodide) and 99mTc (Pertechnetate) in the thyroid of neonates using Cristy-Eckerman and Segars anatomical representations

Using the MIRD formalism, and the Cristy-Eckeman and Segars anthropomorphic representations, the absorbed dose in the thyroid of newborns, was calculated when 123I (iodide) and 99mTc (pertechnetate) are used during the diagnostic procedures. The dose results will allow exploring the dosimetric impact generated by the use of these radiopharmaceutical compounds and the use of two representations. Regardless the radiopharmaceutical compound and the anthropomorphic representation is the thyroid self-dose is the greatest, due to electrons emitted during the 123I and 99mTc radioisotopes. The relative difference in total dose to the newborn thyroid gland using the Cristy-Eckerman and Segars anthropomorphic representations for the compounds 123I(iodide) and 99mTc(pertechnetate) is 1.82%, and 1.33%, respectively. Regardless of the radiopharmaceutical compound, the replacement of Cristy-Eckerman by Segars phantom does not reflect significant changes in the estimated absorbed dose to the newborn thyroid. Regardless of the anthropomorphic representation, the lowest absorbed dose in newborn's thyroid is obtained when using 99mTc (pertechnetate) is used due to the residence times

10B+ZnS(Ag) as an alternative to 3He-based detectors for Radiation Portal Monitors

Typical radiation portal monitor systems, RPM, deployed to detect illicit trafficking of radioactive materials include a set of gamma-ray detectors and neutron detectors. Usually the employed neutron detectors are pressurized 3He-based neutron detectors tubes. Due the shortage of 3He reported since 2009, the amount of 3He available for use in gas proportional counter neutron detectors has become limited, while the demand has significantly increased, especially for homeland security applications. For this reason, many different alternatives are being investigated for its use in RPM systems. The aim of this work is to study a scintillation detector ZnS(Ag) mixed with highly enriched 10B, 10B+ZnS(Ag). Using Monte Carlo methods, MCNPX code, the response of two neutron detectors based on 10B+ZnS(Ag), manufactured by BridgePort Instruments LLC with different geometries, were estimated by calculating the number of 10B(n,α)7 Li reactions for 29 monoenergetic neutron sources. Measurements and models were made, and both detectors were compared. The importance of the distance with respect to the ground was studied. The response with a 252Cf moderated neutron source (0.5 cm lead and 2.5 cm polyethylene) was calculated in order to compare with other studied alternatives in the USA by Pacific National Northwest Laboratory, PNNL. With these results we conclude that neutron detectors using 10B+ZnS(Ag) are an interesting alternative for replacing 3He detectors. From the analysis with MCNPX we propose an improvement in the detector design

Dosimetric evaluation of radiopharmaceuticals in kidneys and uterine wall of a woman with early pregnancy using Stabin/Segars representations

MIRD method with the Stabin/Segars anthropomorphic representations were used to calculate the absorbed doses in kidneys and uterine wall of an early-stage pregnant woman when 99mTc (DTPA), 99mTc (DMSA) and 99mTc (MAG3) are used for renal studies. Stabin and Segars anatomical representations are phantoms that are used in Monte Carlo calculations to determine the SAF, that with the pharmaceutical residence time are used to calculate the absorbed dose, from source organs, on target organs. Concerns about the impact on the absorbed dose due to the use of the three 99mTc-based compounds as well as the use of different phantoms were here treated for the case of a female at early pregnant state. The lowest absorbed dose in the kidneys was obtained with 99mTc (MAG3), and the relative difference of using Stabin and Segars anthropomorphic representations was 2.5%. For bladder and rest of organs the relative difference 14.63%. The lowest absorbed dose by uterine wall was obtained with 99mTc(DMSA), however the relative difference of using Stabin and Segars anthropomorphic representations was 12%

Photon Shielding Features of Quarry Tuff

Cantera is a quarry tuff widely used in the building industry; in this work the shielding features of cantera were determined. The shielding characteristics were calculated using XCOM and MCNP5 codes for 0.03, 0.07, 0.1, 0.3, 0.662, 1, 2, and 3 MeV photons. With XCOM the mass interaction coefficients, and the total mass attenuation coefficients, were calculated. With the MCNP5 code a transmission experiment was modelled using a point-like source located 42 cm apart from a point-like detector. Between the source and the detector, cantera pieces with different thickness, ranging from 0 to 40 cm were included. The collided and uncollided photon fluence, the Kerma in air and the Ambient dose equivalent were estimated. With the uncollided fluence the linear attenuation coefficients were determined and compared with those calculated with XCOM. The linear attenuation coefficient for 0.662 MeV photons was compared with the coefficient measured with a NaI(Tl)-based γ-ray spectrometer and a 137Cs source