Thermal and resonance neutrons generated by various electron and X-ray therapeutic beams from medical linacs installed in polish oncological centers

Thermal neutron detection and track recognition method in reference and out-of-field radiotherapy FLASH electron fields using Timepix3 detectors

Objective.This work presents a method for enhanced detection, imaging, and measurement of the thermal neutron flux.Approach. Measurements were performed in a water tank, while the detector is positioned out-of-field of a 20 MeV ultra-high pulse dose rate electron beam. A semiconductor pixel detector Timepix3 with a silicon sensor partially covered by a6LiF neutron converter was used to measure the flux, spatial, and time characteristics of the neutron field. To provide absolute measurements of thermal neutron flux, the detection efficiency calibration of the detectors was performed in a reference thermal neutron field. Neutron signals are recognized and discriminated against other particles such as gamma rays and x-rays. This is achieved by the resolving power of the pixel detector using machine learning algorithms and high-resolution pattern recognition analysis of the high-energy tracks created by thermal neutron interactions in the converter.Main results. The resulting thermal neutrons equivalent dose was obtained using conversion factor (2.13(10) pSv·cm2) from thermal neutron fluence to thermal neutron equivalent dose obtained by Monte Carlo simulations. The calibrated detectors were used to characterize scattered radiation created by electron beams. The results at 12.0 cm depth in the beam axis inside of the water for a delivered dose per pulse of 1.85 Gy (pulse length of 2.4μs) at the reference depth, showed a contribution of flux of 4.07(8) × 103particles·cm-2·s-1and equivalent dose of 1.73(3) nSv per pulse, which is lower by ∼9 orders of magnitude than the delivered dose.Significance. The presented methodology for in-water measurements and identification of characteristic thermal neutrons tracks serves for the selective quantification of equivalent dose made by thermal neutrons in out-of-field particle therapy.

FOIL ACTIVATION TECHNIQUE-A TOOL FOR THE EVALUATION OF PHOTO-NEUTRON DOSE IN RADIOTHERAPY

Treatment of cancer is carried out using photon beams from high-energy medical linear accelerators. Photo-neutrons are also produced as an unwanted by product in the process of dose delivery to the cancer patients during their radiation treatments. In the present study, photo-neutron dose equivalents (both thermal and fast components) per unit delivered gamma-photon dose were measured at different depths, as function of distances from iso-centre in patient plane, field sizes, wedge angles and at LINAC head for a 15-MV medical linear accelerator model Elekta Precise using multi-foil activation technique. The neutron dose equivalents determined for the above-mentioned parameters were found to be lower (<0.05%) in comparison with the therapeutic photon dose delivered and within the prescribed limits recommended by the national regulatory authority.

Application of therapeutic linear accelerators for the production of radioisotopes used in nuclear medicine

This review paper summarizes the possibilities of the use of therapeutic linear electron accelerators for the production of radioisotopes for nuclear medicine. This work is based on our published results and the thematically similar papers by other authors, directly related to five medical radioisotopes as 99Mo/99mTc, 198Au, 186Re, 188Re, 117mSn, produced using therapeutic linacs. Our unpublished data relating to the issues discussed have also been used here. In the experiments, two types of radiation were included in the analysis of the radioisotope production process, i.e. the therapeutic twenty-megavolt (20 MV) X-rays generated by Varian linacs and neutron radiation contaminating the therapeutic beam. Thus, the debated radioisotopes are produced in the photonuclear reactions and in the neutron ones. Linear therapeutic accelerators do not allow the production of radioisotopes with high specific activities, but the massive targets can be used instead. Thus, the amount of the produced radioisotopes may be increased. Apart from linear accelerators, more and more often, the production of radioisotopes is carried out in small medical cyclotrons. More such cyclotrons are developed, built, and sold commercially than for scientific research. The radioisotopes produced with the use of therapeutic linacs or cyclotrons can be successfully applied in various laboratory tests and in research.

ESTIMATION OF THERMAL & EPITHERMAL NEUTRON FLUX AND GAMMA DOSE DISTRIBUTION IN A MEDICAL CYCLOTRON FACILITY FOR RADIATION PROTECTION PURPOSES USING GOLD FOILS AND GATE 9

The aim of this study is to characterise the neutron flux generated directly behind targets used in medical cyclotrons. The characterisation process aims at determining the feasibility of using the generated neutrons for research purposes in neutron activation analysis. The study was performed by activating gold foils placed directly behind the cyclotron targets. The thermal and epithermal neutron flux were found to be 4.5E+05 ± 8.78E+04 neutrons cm-2 s-1 and 2.13E+06 ± 8.59E+04 neutrons cm-2 s-1, respectively. The flux value is the same order of magnitude listed in the manual produced by the cyclotron manufacturer. The results are encouraging and show high potential for using the cyclotron facility as a thermal neutron source for research purposes. However, it is important radiation protection procedures be followed to ensure the safety of researchers due to the high gamma dose rate measured directly behind the target at 2.46 Sv/h using an OSL chip during the beam on time.

Experimental determination of thermal neutron fluence around Elekta Versa HD linear accelerator for various photon energies

A complex neutron spectrum generated along with a useful photon beam imposes an additional radiation protection risk around medical linear accelerators (linac). The thermal neutron component of this complex neutron spectrum formed during different photon modes of operation of Elekta Versa HD linac has been quantified using Indium foil activation technique. The thermal neutron fluence (Φ _th ) at isocenter for 15 MV, 10 MV and 10 MV FFF beams was found to be 2.45 × 10⁵, 4.35 × 10⁴ and 3.2 × 10⁴ neutrons cm^-2 Gy^-1, respectively. The analysis shows a reduction in the Φ _th as the flattening filter is being taken out from the beam path. A negative correlation in Φ _th with respect to field size has been observed with an average 18% reduction in Φ _th per monitor units as field size changes from 10 cm × 10 cm to 40 cm × 40 cm. For particular field size and photon energy, Φ _th was found to be uniform across the patient plane. From the measured gamma ray spectrum inside the treatment room six major isotopes have been identified which were ¹²²Sb, ¹⁸⁷W, ⁸²Br, ⁵⁶Mn, ²⁴Na and ²⁸Al.

CubBonner - A novel passive neutron area monitor

Responses of a novel passive neutron area monitor, named CubBonner, were calculated. The responses were estimated for sixty monoenergetic neutrons, from 10^-9 to 20 MeV, with the MCNP5 code. The CubBonner is a cubic polyethylene moderator and a gold foil as thermal neutron detector. The ambient dose equivalent response was calculated for three cubes (5″, 8″ and 10" side) with the gold foil at the cube's centers. The moderator cube having the best ambient dose equivalent response was used to estimate the neutron fluence and the ¹⁹⁷Au(n,γ) responses per history. The ambient dose response per unit mass of gold was compared with the response of the Berthold LB 6411 active neutron area monitor, and the response for the (n,γ) reaction in the gold foil was compared with the evaporation photo neutrons produced in linear accelerators for radiotherapy.

Application of the phase-space distribution approach of Monte Carlo for radiation contamination dose estimation from the (n,γ), (γ,n) nuclear reactions and linac leakage photons in the megavoltage radiotherapy facility

Aim

The aim of this study was to characterize the radiation contamination inside and outside the megavoltage radiotherapy room.

Background

Radiation contamination components in the 18 MV linac room are the secondary neutron, prompt gamma ray, electron and linac leakage radiation.

Materials and Methods

An 18 MV linac modeled in a typical bunker employing the MCNPX code of Monte Carlo. For fast calculation, phase-space distribution (PSD) file modeling was applied and the calculations were conducted for the radiation contamination components dose and spectra at 6 locations inside and outside the bunker.

Results

The results showed that the difference of measured and calculated percent depth-dose (PDD) and photo beam-profile (PBP) datasets were lower than acceptable values. At isocenter, the obtained photon dose and neutron fluence were 2.4 × 10^-14 Gy/initial e° and 2.22 × 10^-8 n°/cm², respectively. Then, neutron apparent source strength (Q_N) value was found as 1.34 × 10¹² n°/Gy X at isocenter and the model verified to photon and neutron calculations. A surface at 2 cm below the flattening filter was modeled as phase-space (PS) file for PDD and PBP calculations. Then by use of a spherical cell in the center of the linac target as a PS surface, contaminant radiations dose, fluence and spectra were estimated at 6 locations in a considerably short time, using the registered history of all particles and photons in the 13GB PSD file as primary source in the second step.

Conclusion

Designing the PSD file in MC modeling helps user to solve the problems with complex geometry and physics precisely in a shorter run-time.

MC safe bunker designing for an 18 MV linac with nanoparticles included primary barriers and effect of the nanoparticles on the shielding aspects

Aim

The aim of this study was to design a safe bunker for an 18 MV linac in to configuration; primary barriers made from nanoparticle-containing concrete and pure concrete.

Background

Application of some nanoparticles in the shielding materials has been studied and it was shown that the presence of some nanoparticles improved radiation shielding properties.

Materials and methods

Some percentage of different nanoparticles were modeled by the MCNP5 code of MC in the megavoltage radiotherapy treatment room's primary barriers. Other parts of the designed room, such as secondary barriers and maze door, were modeled as ordinary pure concrete. A safe bunker was designed according to the MC derived spectra at primary and secondary barriers location using a modeled and benchmarked 18 MV linac in free air. Then, the thickness of the required shielding materials for the door and also concrete for the walls and primary barriers were calculated separately.

Results

According to the results, required concrete thickness in primary and secondary barriers was reduced by around 0.8% compared to pure concrete application. Additionally, required lead and BPE decreased by 25% and 15%, respectively, due to primary barriers nanoparticles.

Conclusions

It was concluded that application of some nanoparticles in the shielding materials structures in megavoltage radiotherapy can make the shielding effective.

Neutron and high energy photon fluence estimation in CLINAC using gold activation foils

Aim

The thermal neutron, epithermal neutron and high-energy photon fluence were measured in this work around the Varian 21EX 23 MV CLINAC, which is operated in Albairouni hospital in Damascus, Syria.

Background

Photoneutron measurements around medical CLINAC aim to protect both patients and staff from unwanted radiation.

Materials and methods

Neutron and photon activation techniques were applied using gold foils.

Results

It was found that high-energy photons fluence has practically a constant value in the field size. The thermal and epithermal neutron fluence along ox and oy axes has the same order of magnitude.

Conclusion

Gold foils have been used successfully to measure neutron flux and high-energy photons simultaneously using activation techniques.

Monte Carlo calculation of photo-neutron dose produced by circular cones at 18 MV photon beams

Aim

The aim of this study is to calculate neutron contamination at the presence of circular cones irradiating by 18 MV photons using Monte Carlo code.

Background

Small photon fields are one of the most useful methods in radiotherapy. One of the techniques for shaping small photon beams is applying circular cones made of lead. Using this method in high energy photon due to neutron contamination is a crucial issue.

Materials and methods

Initially, Varian linac producing 18 MV photons was simulated and after validating the code, various circular cones were also simulated. Then, the number of neutrons, neutron equivalent dose and absorbed dose per Gy of photon dose were calculated along the central axis.

Results

Number of neutrons per Gy of photon dose had their maximum value at depth of 2 cm and these values for 5, 10, 15, 20 and 30 mm circular cones were 9.02, 7.76, 7.61, 6.02 and 5.08 (n cm^-2 Gy^-1), respectively. Neutron equivalent doses per Gy of photon dose had their maximum at the surface of the phantom and these values for mentioned collimators were 1.48, 1.33, 1.31, 1.12 and 1.08 (mSv Gy^-1), respectively. Neutron absorbed doses had their maximum at the surface of the phantom and these values for mentioned collimators sizes were 103.74, 99.71, 95.77, 81.46 and 78.20 (μGy/Gy), respectively.

Conclusions

As the field size gets smaller, number of neutrons, equivalent and absorbed dose per Gy of photon increase. Also, neutron equivalent dose and absorbed dose are maximum at the surface of phantom and then these values will be decreased.

Fast, epithermal and thermal photoneutron dosimetry in air and in tissue equivalent phantom for a high-energy X-ray medical accelerator

Photoneutron (PN) dosimetry in fast, epithermal and thermal energy ranges originated from the beam and albedo neutrons in high-energy X-ray medical accelerators is highly important from scientific, technical, radiation protection and medical physics points of view. Detailed dose equivalents in the fast, epithermal and thermal PN energy ranges in air up to 2m as well as at 35 positions from the central axis of 12 cross sections of the phantom at different depths were determined in 18MV X-ray beams of a Siemens ONCOR accelerator. A novel dosimetry method based on polycarbonate track dosimeters (PCTD)/¹⁰B (with/without cadmium cover) was used to determine and separate different PN dose equivalents in air and in a multilayer polyethylene phantom. Dose equivalent distributions of PNs, as originated from the main beam and/or albedo PNs, on cross-plane, in-plane and diagonal axes in 10cm×10cm fields are reported. PN dose equivalent distributions on the 3 axes have their maxima at the isocenter. Epithermal and thermal PN depth dose equivalent distributions in the phantom for different positions studied peak at ∼3cm depth. The neutron dosimeters used for the first time in such studies are highly effective for separating dose equivalents of PNs in the studied energy ranges (beam and/or albedo). The PN dose equivalent data matrix made available in this paper is highly essential for detailed patient dosimetry in general and for estimating secondary cancer risks in particular.

Experimental validation of neutron activation simulation of a varian medical linear accelerator

This work presents a Monte Carlo simulation using the last version of MCNP, v. 6.1.1, of a Varian CLinAc emitting a 15MeV photon beam. The main objective of the work is to estimate the photoneutron production and activated products inside the medical linear accelerator head. To that, the Varian LinAc head was modelled in detail using the manufacturer information, and the model was generated with a CAD software and exported as a mesh to be included in the particle transport simulation. The model includes the transport of photoneutrons generated by primary photons and the (n, γ) reactions which can result in activation products. The validation of this study was done using experimental measures. Activation products have been identified by in situ gamma spectroscopy placed at the jaws exit of the LinAc shortly after termination of a high energy photon beam irradiation. Comparison between experimental and simulation results shows good agreement.

Significant change in the construction of a door to a room with slowed down neutron field by means of commonly used inexpensive protective materials

The detailed analysis of nuclear reactions occurring in materials of the door is presented for the typical construction of an entrance door to a room with a slowed down neutron field. The changes in the construction of the door were determined to reduce effectively the level of neutron and gamma radiation in the vicinity of the door in a room adjoining the neutron field room. Optimisation of the door construction was performed with the use of Monte Carlo calculations (GEANT4). The construction proposed in this paper bases on the commonly used inexpensive protective materials such as borax (13.4 cm), lead (4 cm) and stainless steel (0.1 and 0.5 cm on the side of the neutron field room and of the adjoining room, respectively). The improved construction of the door, worked out in the presented studies, can be an effective protection against neutrons with energies up to 1 MeV.

Doses in sensitive organs during prostate treatment with a 60Co unit

Using thermoluminiscent dosimeters the absorbed dose in the bladder, rectum and thyroid have been evaluated when 200 cGy was applied to the prostate. The treatment was applied with a (60)Co unit. A water phantom was built and thermoluminiscent dosimeters were located in the position where the prostate, bladder, rectum and thyroid are located. The therapeutic beam was applied in 4 irradiations at 0, 90, 180 and 270° with the prostate at the isocenter. The TLDs readouts were used to evaluate the absorbed dose in each organ. The absorbed doses were used to estimate the effective doses and the probability of developing secondary malignacies in thyroid, rectum and bladder.