Photon Dose Rate Research Articles

Established of Ultra-Low Dose Rate Photon Measurement Facility in China Jinping Underground Laboratory

Established of Ultra-Low Dose Rate Photon Measurement Facility in China Jinping Underground Laboratory

A Comparison of Conventional Empirical Formula and MCNPX Code in the Estimations of Photon and Neutron Skyshine Rates for an 18MV Radiotherapy Bunker

Purpose: A physical phenomenon, scattering the radiation by the atmosphere above the room to the points at ground level around the linac treatment room is known as skyshine radiation. This study aimed to estimate photon and neutron skyshine from a linac in a high-energy radiation therapy facility. Materials and Methods: The empirical method of NCRP report 151 and MC simulations were employed to estimate skyshine radiation dose from the 18MV linac photon beam. A linac and its bunker were modeled and skyshine dose equivalent from photons and secondary neutrons were derived and compared in the control room, corridor, sidewalk and, parking. Results: The photon skyshine dose rates calculations by the MC method varied from 0.43 µSv/h at the sidewalk to 6.2 µSv/h at the control room. The ratios of NCRP to MCNP calculations varied from 3.58 for the corridor to 16.14 for the control room. For the neutron skyshine dose rate at distances shorter than 20m, it was found to be 10.4 nSv/h and the ratios of the NCRP to MCNP were 1.26 at the control room and 3.34 at the sidewalk. Conclusion: It was concluded that the empirical method overestimates photon and neutron skyshine dose rates in comparison to the MCNPX code. The refinement of the proposed empirical method of NCRP 151 and application of MC methods are strongly suggested for more reliable calculations of skyshine radiations.

A novel method for the design and selection of neutron-attenuating media

A new methodology for designing neutron shields and selecting moderators using sets of generalized metrics is presented. The metrics were derived from many MCNP®6.2 simulations. Neutron and photon statistics were tracked at the outer surface of a sphere for each of the materials in each simulation to derive neutron energy attenuation coefficients that can accurately track the exponential energy decrement of neutrons in media. Neutron penetration probabilities as well as a normalized measure of the number of collisions were also obtained as a measure of the number of neutrons that penetrate the media and the number of collisions the neutrons undergo, respectively. Finally, the penetrative photon generation probability and their average energy were also obtained, allowing for photon dose rate computations. Attempts at generalizing each of the derived metrics are also discussed, showing the limitations of such simplifications. The methodology allows more streamlined approaches for analyzing complex neutron attenuating media.

Experimental and Computational Dose Rate Evaluation Using SN and Monte Carlo Method for a Packaged 241AmBe Neutron Source

We present a systematic computational dose rate evaluation for a packaged 1.8-Ci 241AmBe source using both Monte Carlo and deterministic approaches, with some experimental measurements for correlation. The 241AmBe source is stored in an extended 55-gal-drum container. Computational dose rate analysis is performed using MCNP6 (Monte Carlo) and PENTRAN (SN) on the Center for High Performance Computing system at the University of Utah. Limited information is available regarding internal drum shielding construction, and a reverse engineering approach is presented here to estimate the dose rate and compare with measured experimental values. Our analysis shows that a deterministic three-dimensional quadrature (SN) and anisotropic scattering (PN) order of S20P2 is sufficient for dose rate calculations of the 241AmBe source with polyethylene surrounding the source as shielding material. Higher quadrature orders, i.e., at least S70 for neutrons and S40 for photons, are needed in the presence of air due to severe streaming effects, and this is dependent upon the distance between the source and measurement locations. With air surrounding the 241AmBe source, the Monte Carlo method is considered to be better for neutron dose calculations while the SN method is considered better for photon dose calculations. Good agreement from both computational verification and experimental validation are observed for the dose “hot spot” in the extended 55-gal drum. The differences noted between the MCNP6/PENTRAN calculations are within 6% for the neutron dose rate and 30% for the photon dose rate. It is observed that more than 95% of the dose is attributed to neutrons. Detailed studies including a literature data validation, PENTRAN SN convergence study, buildup factor analysis, and dose rates with different shielding materials are presented in the narrative.

Dosimetric effects of composition, location and size of tissue heterogeneities on 252Cf neutron brachytherapy

The purpose of this study is to evaluate the effect of tissue heterogeneities on dose distribution in Californium-252 (252Cf) neutron brachytherapy. The effect of location and size of heterogeneity on dose distribution was also evaluated. Neutron and photon dose rate distributions were determined in a water phantom in presence of air, lung, soft tissue and bone heterogeneities using MCNPX code. To benchmark the Monte Carlo simulation of the 252Cf source, air kerma strength (SKN), dose rate constant (ɅN) and radial dose function (gN(r)) were calculated and compared with previously reported data. Results showed a considerable reduction of neutron dose rate (up to 66%) inside heterogeneities, especially in air and bone heterogeneities, while the reduction of total photon dose rate was found less significant (up to 10%). In the presence of a heterogeneity, dose rate, fluence and energy spectrum were significantly different with respect to the homogenous phantom. The contribution of photon dose to the total dose in the presence of air and bone was dominant, compared to the neutron dose, whereas this photon contribution was reduced after passing the heterogeneity. As the bone heterogeneity size was increased from 1 × 1 × 1 cm3 to 1 × 3 × 1 cm3, the total dose and neutron energy fluence decreased of about 50% and 70%, respectively.

Monte carlo analysis of various geometrical blankets in a fusion-fission hybrid reactor

• A dense plasma focus device design as a fusion neutron source. • Investigate the neutron diffusion equation in the cylindrical layer, the spherical layer and the frustum cone layer geometries. • Investigate the lowest probability of a neutron returning from the second blanket to the first blanket in the cylindrical layer, the spherical layer and the frustum cone layer geometries. • Investigate the neutron parameters in the One-Blanket Reactor and Two-Blanket Reactorand the ultimate energy of the hybrid reactor by MCNPX code. • To check the safety of staff and reactor facilities, the dose rate of neutrons and photons in the concrete shield is calculated by MCNPX code. Fusion - fission hybrid reactor consists of two main parts. The first part is the Dense Plasma Focus (DPF) where the fusion process is performed and the second part is the blankets which surround the DPF and the fission process is done in the blankets. The most of the fusion neutrons, near the fusion source, are directed upward, so the blanket over the DPF is designed in forms cylindrical, spherical and frustum cone of volumes 183,347 cm 3 , 136,378 cm 3 and 58,223.2 cm 3 respectively at the same effective multiplication factor. Then the thermal neutron diffusion equation, buckling and neutron parameters are studied. After selecting the first blanket, the second blanket is designed in forms cylindrical, spherical and frustum cone of volumes 4.500 × 10 6 cm 3 , 3.689 × 10 6 cm 3 and 6.441 × 10 5 cm 3 respectively. So the frustum cone is chosen for the first and second blanket. The neutron return probability in the frustum cone from the second blanket to the first blanket is low, so the first blanket's effective multiplication factor does not change and the reactor's total effective multiplication factor is stable. Also, MCNPX code has been used to checkout and select neutron multiplier and reflector cells for the most neutron yield.

The Spectral Measurement of Scattered Radiation From a Clinical Linear Accelerator Using a CZT Detector

The study of the induced radioactivity following radiotherapy with high energy X-rays from medical linear accelerator. Patient equivalent phantom made of Polymethyl methacrylate (PMMA) of 30x30x27 cm size irradiated with 15 MV X-rays from Versa HD medical linear accelerator form Elekta. Induced radioactive and ambient dose rates were measured at 0.25, 0.5 and 1 m from beam center using GR1® spectrometry with Cadmium Zinc Telluride (CZT) detectors having energy resolution less than 2%. Spectrum analysis was performed using MultiSpect software. The measured spectrum showed 511 keV annihilation photons possibly as a result of positron emitter of which most likely candidates are 62 Cu(T1/2: 9.7 min), 64 Cu (T1/2: 12.7 h ) and 57 Ni (T1/2: 35.6 h) and a peak at ≈ 1780 keV that could be attributed 28 Al and 214 Bi radioisotope. Ambient photon dose rates post radiotherapy treatment ranged 660 µGyh -1 at o.5 m to 41 µGyh -1 at 1 m. These values agree well with the results presented in the literature. Keywords: Radiotherapy; Activation Products; Gamma spectrometry; Occupational exposure; Medical Linear Accelerator. DOI: 10.7176/ALST/83-05 Publication date: November 30 th 2020

DUAL-PARTICLE DOSEMETER BASED ON ORGANIC SCINTILLATOR

Traditionally available handheld dosemeters are generally sensitive to only one type of radiation: neutrons or photons. Some dosemeters also rely on very specific attenuation correlations between response and dose, are not scalable in size and multiple dosemeters are required to characterise mixed-particle fields. The research presented here serves as a proof-of-concept for a method to simultaneously measure dose rates from neutrons and photons using a particle discriminating organic scintillation detector without the need for spectral deconvolution. The method was compared with traditional instruments and to simulation. Isotopic photon dose rates measured with this method were within 4% of simulated truth, whereas fission spectrum neutron dose rates were measured within 21%. Measurements of dose rates from both particles agree with simulated truth better than traditional instruments. This new method allows for measurement of dose equivalent from both neutrons and photons with a single instrument and no reliance on spectral deconvolution.

Neutron and Photon Dose Rates in a D-T Neutron Generator Facility: MCNP Simulations and Experiments.

The deuterium-tritium neutron generator is a common neutron source for fast neutron activation analysis. The 14.1 MeV neutrons emitted from a deuterium-tritium neutron generator are difficult to shield due to their strong penetrability and the induced secondary gamma rays in the shield. A rough calculation based on attenuation factors shows that when 14.1 MeV neutrons with a yield of 1 × 10 ns penetrate the designed shielding layers, which consist of a 0.5-m-thick concrete layer and a 0.5-m-thick water layer, the neutron ambient dose equivalent rate is 2.48 μSv h. A geometric model of a neutron shielding room is constructed based on the calculation. Monte Carlo simulations indicate that the highest neutron ambient dose equivalent rate outside the neutron shielding room is 0.73 μSv h, and the neutron ambient dose equivalent rate at the detector position in the shielding room is 2.12 μSv h. The experimental results show that the highest neutron ambient dose equivalent rate outside the neutron shielding room is 1.43 μSv h, and the neutron ambient dose equivalent rate at the detector position inside the shielding corridor is 2.74 μSv h. Comparative investigations show that the experimental results are basically consistent with the results of the Monte Carlo simulations, except for some positions with large proportions of fast neutrons where it is too difficult for the neutron dose equivalent meter to provide reliable values. Moreover, the radiation dose rate outside the designed shielding room is lower than the occupational exposure dose limit, which is in line with the design expectations. Finally, the gamma spectrum at the position of the gamma detectors is measured by a high-purity germanium detector. The analyzed results show that many secondary gamma rays are generated by the interaction of neutrons with the shield materials and detector probe crystals, and some gamma rays are produced from natural background radionuclides such as K, Tl, Bi, Bi, Pb, Pb, and Ac.

A Monte Carlo study of neutron contamination in presence of circular cones during stereotactic radiotherapy with 18 MV photon beams

High-energy photons are being used to treat different kinds of cancer, but it may increase the rate of secondary cancers due to the neutron contamination as well as over exposing of patients and medical staffs in radiation therapy Takam, Bezak, Marcu, and Yeoh, 2011, Radiation Research, 176, 508-520. Due to some difficulties in experimental measurements of neutron contamination, Monte Carlo method is an efficient tool to investigate dose parameters and characteristics in new techniques. The 18-MV photon beam of linac and circular cones have been simulated by MCNP5 code. Various parameters of photon and neutron including mean energy, flux, KERMA, the number of particles crossing a surface at a distance of 100 cm (SSD = 100 cm) as well as the change in photon and neutron spectrum as well as in intensity through the transmission in the circular collimators have been investigated. The results of this study show that the use of a circular collimator decreases neutron dose in the central axis, which is an advantage, but neutron contamination inducing small neutron dose is distributed all over the space. On the surface of phantom, photon dose rate is approximately equal to 3.41E7 (mGy/mA.min) for different collimators, but the neutron dose rate is 1.64E2 (mGy/ mA.min), 2.03E2 (mGy/ mA.min) and 2.52E2 (mGy/mA.min) for diameters of 12, 20 and 40 mm, respectively and it decreases by decreasing the diameter of the collimator. The neutron dose rate decreases from 9.68E7 and 9.74E7 (mGy/min.mA) for open field size 33 cm2 and 55 cm2 to 1.64E2 (mGy/min.mA), 2.02E2 (mGy/min.mA) and 2.52E2 (mGy/min.mA) for collimator diameter of 12 mm, 20 mm and 40 mm. It can be concluded that the use of circular collimators has an advantage of reducing neutron dose in the central axis. It should be mentioned that the off-axis neutron dose surrounding the collimator can be eliminated using an external neutron shield without perturbing the treatment field.

Photon energy response optimization using few-channel spectroscopy dose method for Si-PIN photodetector applied in personal dose equivalent measurements

Si-PIN photodetectors having features such as low cost, small size, low weight, low voltage, and low power consumption are widely used as radiation detectors in electronic personal dosimeters (EPDs). The technical parameters of EPDs based on the Si-PIN photodetectors include photon energy response (PER), angular response, inherent error, and dose rate linearity. Among them, PER is a key parameter for evaluation of EPD measurement accuracy. At present, owing to the limitations of volume, power consumption, and EPD cost, the PER is usually corrected by a combination of single-channel counting techniques and filtering material methods. However, the above-mentioned methods have problems such as poor PER and low measurement accuracy. To solve such problems, in this study, a 1024-channel spectrometry system using a Si-PIN photodetector was developed and full-spectrum measurement in the reference radiation fields was conducted for radiation protection. The measurement results using the few-channel spectroscopy dose method showed that the PER could be controlled within ± 14% and ± 2% under the conditions of two and three energy intervals, respectively, with different channel numbers. The PER measured at 0° angle of radiation incidence meets the − 29% to +67% requirements of IEC 61526:2010. Meanwhile, the channel number and counts-to-dose conversion factors formed in the experiment can be integrated into an EPD.

Assessment of photon energy and dose rate dependence of U-NIPAM polymer gel dosimeter

One of the main drawbacks of NPAM gel dosimeter is its relatively low R2-dose sensitivity. Recently, the sensitivity of NIPAM gel dosimeter has been improved by adding urea and this new formula gel dosimeter was named U-NIPAM (Urea and NIPAM). In the current project, photon energy and dose rate dependence of optimized U-NIPAM gel dosimeter (with 3% urea) was assessed by MRI technique. The gel samples were fabricated under normal atmospheric conditions and after 1 day, they were exposed to photon energies of 6 and 15 MV and dose rates of 300, 400, and 500 cGy/min. Then, MRI responses (R2) of the U-NIPAM gel samples were analyzed at 12, 36, 72, 168 h after the irradiation process. The dosimetric evaluation of optimized U-NIPAM gel dosimeter characteristics at 12 h after the irradiation process resulted to following findings: 1) different photon energies (6 and 15 MV) have significant effect on the R2-dose sensitivity of gel dosimeter in 0–7 Gy dose range (P < 0.05), 2) the R2-dose sensitivity of gel dosimeter does not significantly change with different dose rate (300–500 cGy/min) in 0–7 Gy dose range (P˃0.05), 3) the obtained dose resolution for evaluated various photon energies and dose rates ranges from 0.06 Gy to 0.20 Gy. Moreover, it was found that the R2-dose sensitivity dependence of gel dosimeter on photon energy can vary over post irradiation; nevertheless, the R2-dose sensitivity dependence of dosimeter on dose rate dose not vary over post irradiation time.

PHOTON FIELD OF ~100–200 KEV FOR ENVIRONMENTAL DOSEMETER CALIBRATION

As a reference photon field, several radionuclides have been used frequently, such as 241Am,137Cs and60 Co for calibration. These nuclides provide mono-energy photons for dosemeters covering few tens of keV-MeV. The main energy around 200keV is important for both environmental and medical fields since the former should consider scattering photons and the later should measure photons from X-ray generator. In our previous work, a backscattered layout can provide a uniform photon field spectra and dose rate with an energy of 190keV by using an affordable intensity 137 Cs gamma source. Several other quasi-monoenergetic photon fields in the range of 100-200keV could be obtained by using several available gamma sources. Two calibrated environmental CsI(Tl) survey meters, Horiba PA-1000 and Mr. Gamma A2700, had been measured with the developed backscattered photon field to understand energy-dependent features in order to confirm dosemeter readings. Consequently, both scintillator instruments are sensitive for measurements of the relatively low dose rates at 190keV.

Extension of Monte Carlo code MCS to spent fuel cask shielding analysis

The shielding analysis of the spent fuel dry storage cask TN-32 is carried out using the continuous-energy Monte Carlo neutron- and photon-transport code MCS developed by the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The shielding analysis involves the transport simulation of the decay neutrons and gammas emitted by the spent fuel to obtain the neutron and gamma dose rate values outside of the cask. This deep-penetration problem is solved by means of weight window population control as variance reduction. The TN-32 cask calculations are repeated with the MCNP6 Monte Carlo code for verification purpose and a good agreement within three standard deviations is observed between MCS and MCNP6. The neutron dose rate on the cask surface and 2 m away from the cask can be calculated with MCS, whereas the photon dose rate calculation for positions outside of the cask will be further improved by employing more sophisticated variance reduction techniques.

Method for the differentiation of radiation-induced photocurrent from total measured current in P3HT/PCBM BHJ photodiodes

Thin film radiation-detecting diodes fabricated in the laboratory, such as an organic bulk heterojunction, often contain conductive leads, indium tin oxide traces and metallic interconnects which are exposed to the high-energy photon beam during operation. These components generate extraneous radiation-induced currents, that, if not accounted for, will erroneously be attributed to the detector. In commercial devices, these contributions are mitigated by minimizing the size of these components, an approach that is often not feasible in a research lab. Here we demonstrate a method to measure these extraneous signals, and by subtraction, correct the gross signal to accurately reflect the signal generated in the active volume of the diode itself.The method can effectively correct the extraneous signal.The method showed promise over a range of photon beam energies, dose rates, and field sizes.

Benchmarking of MANCINTAP 2.0, a parallel high-performance tool for neutron activation analysis in complex 4D scenarios

MANCINTAP is a parallel computational tool developed and maintained by Ansaldo Nucleare since 2011 to perform 4D neutron transport, activation and time-resolved dose-rate calculations in complex geometries for CPU-intensive applications. MANCINTAP creates an automated link between the 3D radiation transport code MCNP5—which is used to evaluate both the neutron fluxes for activation calculations and the resulting secondary photon dose rates—and the zero-dimensional activation code Anita2000 by handling crucial processes such as data exchange, determination of material mixtures and generation of cumulative probability distributions. Ansaldo Nucleare recently developed a new version of the tool, named MANCINTAP 2.0. A brief description of the computational tool is given here, with particular emphasis on the key technical choices underlying the project. The main enhancement is here highlighted with respect to the former version, and the results of a benchmarking campaign of MANCINTAP 2.0 against the Monte Carlo code FLUKA are described for two test cases.

The Importance and Clinical Implications of FLASH Ultra-High Dose-Rate Studies for Proton and Heavy Ion Radiotherapy.

The goal of radiation therapy is to provide the highest probability of tumor control while minimizing normal tissue toxicity. Recently, it has been discovered that ultra-high dose rates of ionizing radiation may preferentially spare normal tissue over tumor tissue. This effect, referred to as FLASH radiotherapy, has been observed in various animal models as well as, more recently, in a human patient. This effect may be related to the cell sparing found in vitro at ultra-high dose rates of photons and electrons dating back to the 1960s. Conditions representative of physiologic oxygen were found to be essential for this process to occur. However, there is no conclusive data on whether this effect occurs with protons, as all results to date have been in cells irradiated at ambient oxygen conditions. There have been no ultra-high dose-rate experiments with heavy ions, which would be relevant to the implementation of FLASH to carbon-ion therapy. These basic science results are critical in guiding this rapidly advancing field, since clinical particle therapy machines capable of FLASH dose rates have already been promoted for protons. To help ensure FLASH radiotherapy is reliable and maximally effective, the radiobiology must keep ahead of the clinical implementation to help guide it. In this context, in vitro and in vivo proton and heavy ion experiments involving FLASH dose rates need to be performed to evaluate not only short-term consequences, but also sequelae related to long-term health risks. Critical to these future studies is consideration of relevant oxygen tensions at the time of irradiation, as well as appropriate in silico modeling to assist in understanding the initial physicochemical events.

Update and New Features of NTHU Flight Dose Calculator: A Tool for Estimating Aviation Route Doses and Cumulative Spectra of Cosmic Rays in Atmosphere

This paper presents a major improvement in the NTHU Flight Dose Calculator, which not only extends its capabilities but also introduces three new features: the automatic batch analysis of multiple flights, route dose summation over all scheduled flights, and estimation of the cumulative spectra of cosmic rays along a flight path. Based on the results of extensive FLUKA simulations of atmospheric showers generated by galactic cosmic rays, a global database containing energy spectra and effective dose rates of neutrons, protons, muons, pions, electrons, positrons, photons, and heavy ions at various altitudes ranging from sea level up to 70 km was constructed and parameterized accounting for the dependences of altitude, geomagnetic cutoff rigidity, and solar modulation. The dose calculator was revised to accommodate and utilize the newly expanded database, which enables the estimation of aviation route doses and cumulative spectra of various radiation components by performing a series of interpolations and integration along a path. A program is a user-friendly tool for assessing radiation exposure of air travel due to galactic cosmic rays. The new features of the batch processing and route dose summation greatly facilitate studies of flight route effects and individual dose monitoring of aircrew. The cumulative spectra of various components can provide additional information for in-depth dosimetry analyses, designing experiments for measuring related quantities, and estimating single-event effects in avionics.

Neutron and photon dose rate distributions in linac rooms

Neutron and photon dose rate distributions in linac rooms

Evaluation of Dose Rate and Photon Energy Dependence of Gafchromic EBT3 Film Irradiating with 6 MV and Co-60 Photon Beams.

Gafchromic films are utilized for two-dimensional dose distribution measurements, especially in radiotherapy. In this study, we investigated a close connection between energy and dose rate of Gafchromic EBT3 films irradiating with 6 MV and Co-60 photon beams over a broad dose range. EBT3 films were exposed to 6 MV and Co-60 photon beams using 4 and 2 Gy/min dose rates over a 10–400 cGy dose range. The films were scanned in red, green, and blue channels to obtain the optical density (OD)–dose curves. The OD–dose curves resulted from three-color scans for different photon energies and dose rates were compared by statistical independent t-test. For the radiations of Co-60 and 6 MV photon beams, the highest correlation was obtained between the 2 and 4 Gy/min dose rates with red and green channels, respectively. Moreover, the red channel had a greater OD response per dose value, following the green and blue channels. There was no significant difference between different photon energies’ (Co-60 and 6 MV) and dose rates’ (2 and 4 Gy/min) dependence on OD-dose response of EBT3 films over a broad domain of radiation dose, except for different photon energies in the blue channel. Our results revealed that the OD-dose response of EBT3 films is independent on photon energies (Co-60 and 6 MV) and dose rate (2 and 4 Gy/min) in the evaluated dose range (10–400 cGy). Therefore, the EBT3 films are suitable, consistent, and reliable instruments for dose measurements in radiotherapy.