Analysis of the Bremsstrahlung Photons Flux and the Neutrons Beams during the Operation of the Medical Electron Accelerator

Purpose: To estimate the contribution of the secondary neutron flux to the total radiation flux during the operation of Trilogy linear medical accelerator and Varian’s Clinac 2100 accelerator for assessment of impact on the health of patients and medical personnel. High-energy linear accelerators operating at energies higher than 8 MeV generate neutron fluxes when interacting with accelerator elements and with structural materials of the room for treating patients. Neutrons can form at the accelerator head (target, collimators, smoothing filter, etc.), the procedure room, and directly in the patient’s body. Because of the high radiobiological hazard of neutron radiation, its contribution to the total beam flux, even at a level of few percent, substantially increases the dose received by the patient. Material and methods: Secondary neutron fluxes were investigated during the process of the linear medical accelerators Trilogy and Clinac 2100 of Varian operation by the photoactivation method using (γ, n) and (n, γ) reactions on the detection target of natural 181Ta. In addition, measurements of neutron spectra were carried out directly in the room during the operation of a medical accelerator using a spectrometer-dosimeter SDMF-1608. Results: It was determined that the neutron flux on the tantalum target is 16 % of the gamma-ray flux on the same target when the accelerator is operated with a 18 MeV bremsstrahlung energy and 5 % when the accelerator is operated with a 20 MeV excluding thermal neutrons. Conclusion: Finally, it may be noted that, taking into account the coefficient of relative biological efficiency (RBE) of neutron radiation for neutrons with energies of 0.1–200 keV equal to 10 compared with the RBE coefficient for gamma quanta (equal to 1), even preliminary analysis demonstrates significant underestimation of the contribution of neutrons dose to the total dose received by the patient in radiation therapy using bremsstrahlung of 18 and 20 MeV.

Download Full-text

The neutron dose equivalent around high energy medical electron linear accelerators

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp1403207p ◽

2014 ◽

Vol 29 (3) ◽

pp. 207-212 ◽

Cited By ~ 2

Author(s):

Marina Poje ◽

Ana Ivkovic ◽

Slaven Jurkovic ◽

Gordana Zauhar ◽

Branko Vukovic ◽

...

Keyword(s):

Neutron Radiation ◽

Medical Personnel ◽

Dose Dependence ◽

High Energy ◽

Neutron Dose ◽

Dose Equivalent ◽

Linear Accelerators ◽

Dose Rates ◽

Neutron Dose Equivalent ◽

Conservative Estimation

The measurement of neutron dose equivalent was made in four dual energy linear accelerator rooms. Two of the rooms were reconstructed after decommissioning of 60Co units, so the main limitation was the space. The measurements were performed by a nuclear track etched detectors LR-115 associated with the converter (radiator) that consist of 10B and with the active neutron detector Thermo BIOREM FHT 742. The detectors were set at several locations to evaluate the neutron ambient dose equivalent and/or neutron dose rate to which medical personnel could be exposed. Also, the neutron dose dependence on collimator aperture was analyzed. The obtained neutron dose rates outside the accelerator rooms were several times smaller than the neutron dose rates inside the accelerator rooms. Nevertheless, the measured neutron dose equivalent was not negligible from the aspect of the personal dosimetry with almost 2 mSv a year per person in the areas occupied by staff (conservative estimation). In rooms with 15 MV accelerators, the neutron exposure to the personnel was significantly lower than in the rooms having 18 MV accelerators installed. It was even more pronounced in the room reconstructed after the 60Co decommissioning. This study confirms that shielding from the neutron radiation should be considered when building vaults for high energy linear accelerators, especially when the space constraints exist.

Download Full-text

DOSE AND GAMMA-RAY SPECTRA FROM NEUTRON-INDUCED RADIOACTIVITY IN MEDICAL LINEAR ACCELERATORS FOLLOWING HIGH-ENERGY TOTAL BODY IRRADIATION

Radiation Protection Dosimetry ◽

10.1093/rpd/ncv480 ◽

2015 ◽

Vol 172 (4) ◽

pp. 327-332 ◽

Cited By ~ 5

Author(s):

S. Keehan ◽

M. L. Taylor ◽

R. L. Smith ◽

L. Dunn ◽

T. Kron ◽

...

Keyword(s):

Total Body Irradiation ◽

Gamma Ray ◽

High Energy ◽

Linear Accelerators ◽

Medical Linear Accelerators ◽

Total Body ◽

Induced Radioactivity

Download Full-text

Additional dose assessment from the activation of high-energy linear accelerators used in radiation therapy

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp0802058a ◽

2008 ◽

Vol 23 (2) ◽

pp. 58-64 ◽

Cited By ~ 4

Author(s):

Embarka Ateia ◽

Olivera Ciraj-Bjelac ◽

Milojko Kovacevic ◽

Petar Belicev ◽

Bratislav Cvetkovic ◽

...

Keyword(s):

Spectral Analysis ◽

Dose Rate ◽

Linear Accelerator ◽

Gamma Ray ◽

High Energy ◽

Dose Assessment ◽

Whole Body ◽

Gamma Ray Spectrometry ◽

Additional Dose ◽

Linear Accelerators

It is well known that medical linear accelerators generate activation products when operated above certain electron (photon) energies. The aim of the present work is to assess the activation behavior of a medium-energy radiotherapy linear accelerator by applying in situ gamma-ray spectrometry and dose measurements, and to estimate the additional dose to radiotherapy staff on the basis of these results. Spectral analysis was performed parallel to dose rate measurements in the isocenter of the linear accelerator, immediately after the termination of irradiation. The following radioisotopes were detected by spectral analysis: 28Al, 62Cu, 56Mn, 64Cu, 187W, and 57Ni. The short-lived isotopes such as 28Al and 62Cu are the most important factors of the clinical routine, while the contribution to the radiation dose of medium-lived isotopes such as 56Mn, 57Ni, 64Cu, and 187W increases during the working day. Measured dose rates at the isocenter ranged from 2.2 ?Sv/h to 10 ?Sv/h in various measuring points of interest for the members of the radiotherapy staff. Within the period of 10 minutes, the dose rate decreased to values of 0.8 ?Sv/h. According to actual workloads in radiotherapy departments, a realistic exposure scenario was set, resulting in a maximal additional annual whole body dose to the radiotherapy staff of about 3.5 mSv.

Download Full-text

Investigation of thermal neutron detection capability of a CdZnTe detector in a mixed gamma-neutron radiation field

Nukleonika ◽

10.2478/nuka-2018-0007 ◽

2018 ◽

Vol 63 (3) ◽

pp. 59-64

Author(s):

Haluk Yücel ◽

R. Bora Narttürk ◽

Senem Zümrüt ◽

Gizem Gedik ◽

Mustafa Karadag

Keyword(s):

Neutron Flux ◽

Thermal Neutron ◽

Radiation Field ◽

Epithermal Neutron ◽

Gamma Ray ◽

Neutron Detection ◽

Neutron Radiation ◽

Detection Sensitivity ◽

Shielding Effect ◽

Cdznte Detector

Abstract The aim of this study was to investigate the thermal neutron measurement capability of a CdZnTe detector irradiated in a mixed gamma-neutron radiation field. A CdZnTe detector was irradiated in one of the irradiation tubes of a 241Am-Be source unit to determine the sensitivity factors of the detector in terms of peak count rate (counts per second [cps]) per neutron flux (in square centimeters per second) [cps/neutron·cm−2·s−1]. The CdZnTe detector was covered in a 1-mm-thick cadmium (Cd) cylindrical box to completely absorb incoming thermal neutrons via 113Cd(n,γ) capture reactions. To achieve, this Cd-covered CdZnTe detector was placed in a well-thermalized neutron field (f-ratio = 50.9 ± 1.3) in the irradiation tube of the 241Am-Be neutron source. The gamma-ray spectra were acquired, and the most intense gamma-ray peak at 558 keV (0.74 γ/n) was evaluated to estimate the thermal neutron flux. The epithermal component was also estimated from the bare CdZnTe detector irradiation because the epithermal neutron cutoff energy is about 0.55 eV at the 1-mm-thick Cd filter. A high-density polyethylene moderating cylinder box can also be fitted into the Cd filter box to enhance thermal sensitivity because of moderation of the epithermal neutron component. Neutron detection sensitivity was determined from the measured count rates from the 558 keV photopeak, using the measured neutron fluxes at different irradiation positions. The results indicate that the CdZnTe detector can serve as a neutron detector in mixed gamma-neutron radiation fields, such as reactors, neutron generators, linear accelerators, and isotopic neutron sources. New thermal neutron filters, such as Gd and Tb foils, can be tested instead of the Cd filter due to its serious gamma-shielding effect.

Download Full-text

Influence of head cover on the neutron dose equivalent in Monte Carlo simulations of high energy medical linear accelerator

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp1802217b ◽

2018 ◽

Vol 33 (2) ◽

pp. 217-222

Author(s):

Hrvoje Brkic ◽

Mladen Kasabasic ◽

Ana Ivkovic ◽

Dejan Agic ◽

Ivana Krpan ◽

...

Keyword(s):

Monte Carlo ◽

Monte Carlo Simulations ◽

Medical Personnel ◽

High Energy ◽

Medical Professional ◽

Linear Accelerators ◽

Flux Change ◽

Place Of Origin ◽

The One ◽

Head Cover

Neutron contamination of radiotherapeutic photon beam occurs when energies higher than 10 MeV are used in radiotherapy. To correctly assess the neutron doses that medical personnel and patients receive, it is highly important to know the spectra of the produced photoneutrons. One of the most common ways to determine such spectrum is to perform Monte Carlo simulations of the accelerator. Major issue in the Monte Carlo modelling is that the manufacturers often does not provide full specifications of the accelerators head, so some parts of the head are omitted from the simulation. Within this paper we present a model that includes head cover compared to the one where it is omitted, as it can often be found in the references. Neutron fluxes, spectra, mean energies and place of origin are compared in isocenter, at the point 1 m above target and the point 1 m aside from the target, in both models. In all the considered planes the flux change was found to be more than 20 %, with a significantly change in neutron energy, what is also important in neutron dosimetry. Ignoring the head cover in the Monte Carlo modelling of the high energy electron linear accelerators in radiotherapy, will introduce a large uncertainty of neutron doses assessing a patient, or a medical professional.

Download Full-text

Neutron Detection in Mixed Neutron-Gamma Fields with Common NaI(Tl) Detectors

10.36227/techrxiv.17317175 ◽

2021 ◽

Author(s):

Guntram Pausch ◽

Achim Kreuels ◽

Falko Scherwinski ◽

Yong Kong ◽

Mathias Küster ◽

...

Keyword(s):

Neutron Flux ◽

Thermal Neutron ◽

Delay Time ◽

Thermal Neutron Flux ◽

Gamma Ray ◽

Neutron Detection ◽

High Energy ◽

Detection Capabilities ◽

Co Doped

<p>Searching digitized detector signals for piled-up delayed components with distinct energy and delay time signatures is a smart method to provide common NaI(Tl) detectors with additional neutron detection capabilities at no extra cost. This technique nicely complements the idea of neutron detection by analyzing events with high energy depositions above the range of common gamma-ray energies. In combination, both approaches can provide half of the neutron sensitivity offered by a commercial <sup>6</sup>Li co-doped NaI(Tl) (NaIL™) scintillator of the same size, at the price of higher and load-dependent background contributions. Delayed-coincidence techniques are most suitable for neutron monitoring or long-term measurements, where the statistics of the acquired delay-time distributions allows separate fitting of the effect and background contributions. In this case, the thermal neutron flux can be quantified in parallel to gamma-ray spectroscopy at overall detector loads exceeding 10 kcps.</p>

Download Full-text