Dose Rate Calculations Research Articles

OPERATION OF AN INDUSTRIAL ELECTRON ACCELERATOR IN A DOUBLE-BEAM e,X-MODE

Analysis of mixed e,X-radiation formation in output devices of an industrial electron accelerator is conducted. The possibility is demonstrated to obtain an extra radiation channel on the basis of a practically free source of Xrays simultaneously with the main channel of product processing with electron beam. The conditions of production of the secondary radiation in the state of electronic equilibrium at product treatment with scanning electron beam in the main radiation channel are studied by means of computer simulation. The dependence of spatial radiant characteristics of the X-ray radiation on the spectrum of a primary electron beam and surface density of a treated load has been established. For an industrial accelerator LU-10 of NSC KIPT, the regime of object processing in the extra radiation channel is examined. The results of calculation of the X-ray dose rate and its spatial distribution are in good agreement with the experimental data. The comparative capacity of both radiation channels of the plant is analyzed. The extra radiation source can be used for the execution of non-commercial programs like sanitation of cultural artefacts.

Estimates of relative beta radiation doses on central and lateral axes of ruthenium/rhodium COB-type plaque used in eye brachytherapy

The 106Ru/106Rh COB-type plaque has a cut-out section that makes it suitable to be used in eye brachytherapy to treat tumours close to the optical nerve. Nevertheless, this asymmetry makes measurements and calculations of dose rates around this kind of beta applicator more difficult to perform. In this work we present a analytical and numerical method to evaluate the relative dose rates along the central axis of the COB-type plaque and a comparison is made with a result found in literature obtained by means of Monte Carlo simulation.

Comparison of deterministic and stochastic methods for external gamma dose rate calculation in the decommissioning of nuclear power plants

Calculations for individual scenarios are carried out for optimization of radiation protection, waste management, decommissioning planning, management of disposal facilities, and others. Comparative study of various calculation codes for different geometries with multilayer shields was performed. Reference geometries are models of the most frequently used waste packages in the decommissioning of nuclear power plants (NPP) in Slovakia. As representatives of deterministic calculation codes were chosen Visiplan, Microshield and Mercurad. Stochastic codes were represented by MCNP code. The impact of different build-up factor (BUF) selection in deterministic codes was evaluated. Experimental measurements with different multilayer shields were performed and compared to the results obtained by MCNP calculations. The percentage deviation between experiment and MCNP calculations was ±20% for every shielding configuration. Analysis of the differences between results was made. Recommendations for choosing of the most suitable calculation code for given application are formulated in this study.

Benchmark analysis of coupled neutron/decay-gamma transport method NASCA for shutdown dose rate calculation

The NASCA method was developed with coupled neutron/decay-gamma transport scheme for the shutdown dose rate (SDDR) estimation in fusion device. Recently it has been upgraded for the purpose to quantify the dominant nuclides, which results in the production of radionuclides due to the neutron irradiation. In this paper, the upgrading work was put forward through planning to change the flow of the activation process of material. The linear factor between mono-energetic neutron flux and decay gamma intensity has been calculated on each source nuclide substituting compound material in previous NASCA. By using this method, the source nuclides can be singled out according to their contribution to the final dose rate results. The upgraded NASCA method has been implemented within cosRMC particle transport code and validated through ITER shutdown dose rate benchmark analysis. The results have been compared with other R2S codes. It showed the new function in NASCA is available to decide the dominated source nuclide contributors as well as to calculate shutdown dose rate accurately.

Ray-tracing simulation of the radiation dose distribution on the surface of the spherical phantom of the MATROSHKA-R experiment onboard the ISS

Space radiation is one of the main concerns for human space flights. The prediction of the radiation dose for the actual spacecraft geometry is very important for the planning of long-duration missions. We present a numerical method for the fast calculation of the radiation dose rate during a space flight. We demonstrate its application for dose calculations during the first and the second sessions of the MATROSHKA-R space experiment with a spherical tissue-equivalent phantom. The main advantage of the method is the short simulation time, so it can be applied for urgent radiation dose calculations for low-Earth orbit space missions.The method uses depth-dose curve and shield-and-composition distribution functions to calculate a radiation dose at the point of interest. The spacecraft geometry is processed into a shield-and-composition distribution function using a ray-tracing method. Depth-dose curves are calculated using the GEANT4 Monte-Carlo code (version 10.00.P02) for a double-layer aluminum-water shielding. Aluminum-water shielding is a good approximation of the real geometry, as water is a good equivalent for biological tissues, and aluminum is the major material of spacecraft bodies.The method is applied to model the dose distribution on the surface of the spherical phantom in the MATROSHKA-R space experiment. The experiment has been carried out onboard the ISS from 2004 to the present. The absorbed dose was determined in 32 points on the phantom's surface. We find a good agreement between the data obtained in the experiment and our calculation results. The simulation method is thus applicable for future radiation dose predictions for low-Earth orbit missions and experiments.

Progress in nuclear analyses of the ITER TBM Port Plug with Dummy TBMs

To achieve the validation and testing of tritium breeding blanket concepts, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are tested in three equatorial ports of the ITER tokamak. The TBM Port Plugs consist of Frame and TBM-Sets, which will be tested sequentially, each one tailored to the specific plasma operation scenario and test objective. In case a TBM-Set is not available, it can be replaced by a Dummy TBM at any time. TBM Frame and Dummy TBM are made of stainless steel and also need to meet ITER requirements on shielding, heat removal and safety, including activation. The maintenance operations for the TBM Port Plug include various activities in the Port Cell as well in the Hot Cell. In order to assess the requirements, a detailed MCNP model of the TBM Port Plug (PP), the Pipe Forest (PF) inside the Port Interspace (PI) and the Bioshield Plug (BP) was developed and integrated into the ITER neutronics “C-Model”.This analysis concerns the neutronic behaviour of the TBM PP with two Dummy TBMs by means of activation and shut-down dose rate (SDDR) calculations. The analysis includes the evaluation of the radiation fields around the TBM PP inside the ITER tokamak at several cooling times. Two configuration scenarios have been considered, one with a PF and PF structure inside the PI and one without this structure. The results show that the contribution to the SDDR from the TBM PP at 12 days after reactor shut down is small compared to the contribution originating from the PF structure.

A fast gamma-ray dose rate assessment method for complex geometries based on stylized model reconstruction

Abstract A fast gamma-ray dose rate assessment method for complex geometries based on stylized model reconstruction and point-kernel method is proposed in this paper. The complex three-dimensional (3D) geometries are imported as a 3DS format file from 3dsMax software with material and radiometric attributes. Based on 3D stylized model reconstruction of solid mesh, the 3D-geometrical solids are automatically converted into stylized models. In point-kernel calculation, the stylized source models are divided into point kernels and the mean free paths (mfp) are calculated by the intersections between shield stylized models and tracing ray. Compared with MCNP, the proposed method can implement complex 3D geometries visually, and the dose rate calculation is accurate and fast.

The radiation exposure of fish in the period of the Techa river peak contamination

Waterborne radioactive releases into the Techa River from the Mayak Production Association in Russia during 1949–1956 resulted in downstream contamination of the river ecosystem. The discharged liquid waste contained both short-lived isotopes (95Zr, 95Nb, 103,106Ru, 141,144Ce, 91Y, 89Sr and 140Ba with half-life from 3 days to 1.02 years) and the long-lived 90Sr and 137Cs (half-life - 28.79 y and 30.07 y, respectively). Even now, when two half-lives of 90Sr and 137Cs have passed, the contamination in the upper river region (about 70 km from the source of releases) is still relatively high. Current anthropogenic dose rates calculated for the fish of the Techa River depend on the distance along the stream and decrease from 150 to 3 μGy day−1. Radiation exposure of fish is expected to have been much higher at the time of the releases. The aim of the study was to evaluate the dose rates for the most common fish species of the river, viz., roach (Rutilus rutilus), perch (Perca fluviatilis) and pike (Esox lucius), in the period of peak contamination of the upper reaches of the Techa River from 1950 to 1951. To achieve this objective, calculation of both internal and external dose rates was performed. For dose rate calculation, the contamination of the river compartments was modeled, body-size dependent dose coefficients were evaluated, morphometric data were analyzed. Maximum dose rates were obtained for roach; minimum – for pike, it depends on fish lifestyle (time spent at the bottom). In the period before September 1950, fish of the upper reaches are assessed to have been exposed to dose rates exceeding the screening level equal to 240 μGy day−1. From September 1950 up to the end of 1952 the fish dose rates along the Techa River were found to be close to the UNSCEAR threshold equal to 9.6 × 103 μGy day−1 or even much more higher (up to 1.9 × 105 μGy day−1). Extremely high historical dose rates did not lead to the difference in fish size and fish growth rate currently observed in the Techa River and in the comparison waterbody (the Miass River). Discussion includes the description of radiation effects observed currently in the river fish. Today the effects observed in hematopoietic system may be the consequence of radiation exposure of fish over several generations. For example, long term dwelling of fish in the radioactively contaminated environment leads to their adaptation to chronic radiation exposure. At the same time, an increase their sensitivity and decrease their adaptive capacity to respond to other stress factors can be observed.

Розрахунок ризику перевищення дозових лімітів на елементи екосистеми за допомогою ERICA tool

The possibilities of using the ERICA tool software to calculate the dose load on the ecosystem were analyzed. The expediency of using the software to optimize the sampling process in ecosystems with different levels of pollution has been determined. Within the framework of the work, the specific activity of samples of soil and blueberry plants, which were collected on the territory of the Chernobyl exclusion zone with different levels of contamination, was measured. It is determined that the highest contamination, and, consequently, the highest dose of external and internal iradiation in the ecosystem is formed on the territory of the TPRAW Red Forest. Keywords: ionizing radiation, radionuclide contamination, calculation of dose rate

Calculation of absorbed dose rate in the lungs, ribs and skin by the mammosite applicator in breast cancer brachytherapy with MCNP

Introduction: Breast cancer is the most common cancer among women. Brachytherapy is one of the ways to treat breast cancer. In brachytherapy treatments, high-dose sources are used in interstitial placement. Here, Iridium-192 is used. This source emits gamma rays with energies ranging from 136 kV to 1060 kV electron volts. The activity of the used spring is 10 Ci. The simulated source here is a cylinder with a length of 0.5 and a radius of 0.017 cm. In this paper, the modeling of a common breast cancer treatment applicator called MammoSite is done by the Monte Carlo simulation code. Then, calculate the amount of doses in the brachytherapy of the breast, target volume (tumor) and the dose rate of the organs at risk of radiation, including lungs, ribs and skin. Materials and Methods: MCNPX: The MCNPX code: is the based on the Monte Carlo method. The code, with input file information and the use of the cross-section library, solves the problem and produces results in an output file. MIRD phantom: To calculate the dose, a model of the human body is required as a phantom. All organs of the human body are characterized by a detailed description for the calculation of the absorbed dose in this phantom. Mammosite applicator: This device has a catheter that can be attached to the high dose rate afterloading machine that connect to the source. The end of catheter is surrounded by a balloon that can be filled with normal saline or water. The balloon is placed in the lumpectomy cavity where the 34Gy can reach the treatment site. The most common size of this balloon is 4-5 cm and 5-6 cm, which has a volume between 34-113 cm cubic meters. Results: The absorbed dose by the organs of the skin, ribs and target volume (tumor) is constant when the breast is treated right or left, dose rate of: target volume= 0/7, ribs=0/026, skin=0/009 Gy/min, but when the breast is treated right, the absorbed dose in: right lung=0/045, right breast=0/726, left lung=0/005, left breast=0/006 Gy/min and when the breast is treated left, the absorbed dose in: left lung=0/044, left breast=0/726, right lung=0/006, right breast=0/0068 Gy/min. Conclusion: As expected, absorbed dose in the target volume and the treated breast was the highest and after that the lung received the same significant dose. Indeed, the higher the organ absorbed dose to the target dose rate is better. When the right breast was treated, the right lungs, ribs, and skin had the highest absorbed dose to the dose received by the tumor And when the left breast was treated, the left lung, ribs and skin had the highest absorbed dose to the dose received by the target valume or tumor.

A very personal, 35 years long journey in ESR dating

This paper describes a personal journey in ESR dating and gives some insights of the progress made over 35 years in ESR dating of tooth enamel. When I started, samples were irradiated with four additive dose steps of 5, 10, 15 and 20 krad and the linear extrapolation yielded the dose value. An assumed dose rate of 100 mrad/a yielded the age of the sample. Since then I learned that one actually has to measure the dose rate. I also learned that the dose response can be described by a single saturating function, or perhaps two, and that errors can be appropriately calculated. Rather than having a single anisotropic CO2− radical, tooth enamel consists of two anisotropic CO2− radicals and two isotropic CO2− radicals, one stable, one unstable. Also, the dose response shows spatial differences in the production of these different radicals in response to beta and gamma rays. Then there is the problem of uranium uptake over time. We solved this by combining ESR and U-series analyses. We can even measure appropriate beta attenuation factors including sequential laser ablation U-series analyses. Even better, all analyses can be carried out on small enamel fragments, which can be glued back into their original position with little to no visible damage. Still, when using the most sophisticated dose rate calculations, they're often close to 1000 μGy a−1.

Dose Rate Evaluation for the ES-3100 Package with HEU Content Using MCNP, ADVANTG, Monaco, and MAVRIC

This paper presents a comparative study of dose rate calculations for the ES-3100 package with highly enriched uranium (HEU) content for different source configurations using the following computer codes: MCNP, Automated Variance Reduction Generator (ADVANTG)/MCNP, Monaco, and Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC). The Model ES-3100 package was developed at the Y-12 National Security Complex for domestic and international transportation of Type B fissile radioactive material. In this study, six different source configurations (i.e., solid cylinder, cylindrical hemishell, cylindrical shell, rectangular plate, cylindrical rod, and cylindrical segment form) having 36 kg of HEU metal inside the package containment vessel (based on configurations in the ES-3100/HEU safety analysis report for packaging) are evaluated. Dose rates at 1 mm and 1 m from the package surfaces are calculated for these different source configurations. The MCNP and Monaco cases are run without any biasing options to accelerate the convergence. The Consistent Adjoint Driven Importance Sampling and the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) methods developed at the Oak Ridge National Laboratory are implemented in the ADVANTG/MCNP and MAVRIC codes to accelerate the convergence. ADVANTG generates variance reduction parameters using the Denovo code, and MCNP is used with the variance reduction parameters to accelerate the convergence. MAVRIC uses the Denovo code to construct an importance map and a biased source distribution that are supplied to Monaco to accelerate the Monte Carlo simulation. The FW-CADIS option in ADVANTG and MAVRIC is used to accelerate the convergence in this study. The accelerated convergence cases (ADVANTG/MCNP and MAVRIC) are about 100 times faster with 100 times less particle simulation than those cases run without biasing options (analog MCNP and analog Monaco). The MCNP, ADVANTG/MCNP, Monaco, and MAVRIC calculated dose rates at 1 mm and 1 m from the package surfaces for the different source configurations are compared and are found to be in general agreement.

Simulation calculation of cosmic radiationdose at latitude 0°

Cosmic radiation is a type of natural radiation that comes from outer space. Above sea level, the dose rate of cosmic radiation is 0.03 μSv and will increase rapidly with increase altitude, even reach 2.38 μSv at 10 km altitude. Cosmic radiation makes the flight crew receive higher radiation than the general public.Simulation of the calculation of cosmic radiation dose rates based on altitude at latitude 0° have been done. This simulation using radiation energy data at several points above sea level. This data is then interpolated to obtain continuous radiation energy data from a height of 0 to 16 km with an increase of 0.1 km. The radiation dose rate calculation is carried out using a simple dosimetry equation. The results have been compared with several studies that have been conducted where the deviation value is below to 40% for a height of 9 to 16 km with a difference in the dose rate of 10−7. While at an altitude below 9 km the deviation value is still quite high so a correction factor is needed.

Shutdown dose rate calculation along the ITER IVVS port

The ITER In-Vessel Viewing System (IVVS) consists of six identical units located at the B1 level of the Tokamak complex, at lower ports 3, 5, 9, 11, 15 and 17. It can be used for visual inspection of the plasma chamber in between plasma pulses. This work is devoted to the nuclear analyses performed for the latest design of the IVVS, to assess the shutdown dose rate (SDDR) in the port cell (PC) and the gallery at 1 day after shutdown in order to verify the design limit in the maintenance area. This analysis was performed through a combined use of the R2S method, for the activation of components from the vacuum vessel VV up to the bioshield, and Advanced D1S method for activation of components from the bioshield up to the gallery. Several MCNP calculations have been performed to transport neutrons and decay gammas up to the bioshield, then, by means of secondary sources, the radiation field has been extended in the PC up to the gallery. This strategy allowed to provide high resolution (voxel size 10 × 10 × 10 cm3) maps of shutdown dose rate in the PC with a statistical uncertainty within ±1% and in the gallery with a statistical uncertainty within ±20%. The consistency of the SDDR results obtained by the two codes at the interface highlighted the reliability of the coupled techniques used in the present analyses. Moreover it is assessed that the design target of 10 μSv/h in the PC is satisfied for the IVVS port.

Calculation of photon attenuation coefficient and dose rate in concrete with the addition of SiO2 and MnFe2O4 nanoparticles using MCNPX code and comparison with experimental results

One of the most important safety features of nuclear facilities is the shielding material used to protect the operating personnel from radiation exposure. The most common materials used in radiation shielding are concretes. In this study, a Monte Carlo N-Particle eXtended code is used to calculate the gamma-ray attenuation coefficients and dose rates for a new concrete material composed of MnFe2O4 nanoparticles, which is then compared with the theoretical and experimental results obtained for a SiO2 nanoparticle concrete material. According to the results, the average relative differences between the simulations and the theoretical and experimental results for the linear attenuation coefficient (μ) in the SiO2 nanoparticle materials are 6.4% and 5.5%, respectively. By increasing the SiO2 content up to 1.5% and the temperature of MnFe2O4 up to 673 K, μ is increased for all energies. In addition, the photon dose rate decreases up to 9.2% and 3.7% for MnFe2O4 and SiO2 for gamma-ray energies of 0.511 and 1.274 MeV, respectively. Therefore, it was concluded that the addition of SiO2 and MnFe2O4 nanoparticles to concrete improves its nuclear properties and could lead to it being more useful in radiation shielding.

Solar Cosmic Ray Dose Rate Assessments During GLE 72 Using MIRA and PANDOCA

AbstractGround level enhancement (GLE) 72, which occurred 10 September 2017, is the most recent of two solar particle‐induced enhancements in ground level measurements of cosmic radiation secondary neutrons in solar cycle 24. GLEs have been unusually rare in this solar cycle. GLEs can significantly increase ionizing radiation dose rates at aviation altitudes for hours to days, leading to concern among crewmembers. Real‐time monitoring and preliminary evaluation of solar proton events, including GLEs, in regard to effective dose rates at aviation altitudes has been ongoing since the U.S. Federal Aviation Administration began operating its Solar Radiation Alert System (SRAS) in 2002. Since then, SRAS has been revised multiple times. In this report, model calculations of dose rates during GLE 72 from Maps of Ionizing Radiation in the Atmosphere (MIRA), the latest SRAS software based on CARI‐7A, are compared with those from the model Professional Aviation Dose Calculator (PANDOCA) developed by the German Aerospace Center (Deutsches Zentrum für Luft‐ und Raumfahrt). At very low cutoff rigidities model calculations agree within 40% and indicate no significant increase in radiation exposures at commercial aviation altitudes. The larger than expected differences at very low cutoff rigidities indicate Geostationary Orbiting Environmental Satellite particle flux data alone that are insufficient to produce consistent solar particle dose estimates.

2D modelling: A Monte Carlo approach for assessing heterogeneous beta dose rates in luminescence and ESR dating: Paper ΙΙ, application to igneous rocks

Accurate evaluations of beta dose rates are crucial in luminescence and electron spin resonance (ESR) thermochronology. Most igneous rocks are heterogeneous and have complex mineralogical structures that render them unsuitable for the conventional methods of beta dose rate calculations based on infinite matrix assumptions. The recently developed software DosiVox-2D provides a Monte-Carlo approach for calculating beta dose rates in realistic heterogeneous geometries. In this paper, we present 2D simulations of uniform and layered igneous rocks. For the modelling, mineral distribution maps were obtained by QEM-EDS (quantitative evaluation of minerals using energy dispersive spectroscopy) and radionuclide concentrations in minerals were determined by laser ablation (LA) ICP-MS. The results show that the skewness of beta dose rate distributions in quartz reduces as the K concentration in a rock increases, and the estimation of the beta dose rate can be critically influenced by the U and Th concentrations and distributions. • QEMSCAN ® provides realistic 2D mineral mapping for heterogeneous rocks. • A single set of radionuclide concentrations cannot be representative for some U and Th enriched minerals. • Selections of representative slices, mapping resolution and size are essential to obtain reliable beta dose rate. • Beta dose rate calculation can be critically influenced by the U and Th concentrations and distributions. • DosiVox-2D enables to estimate beta dose rate in single-grain level.

2D modelling: A Monte Carlo approach for assessing heterogeneous beta dose rate in luminescence and ESR dating: Paper I, theory and verification

Many samples used for luminescence and ESR dating show complex mineralogical structures, resulting in heterogeneous beta dose rates. Ideally, the beta dose rate distribution could be reconstructed using softwares like DosiVox, which can both upload 3D-scans recorded with X-ray tomography and distributions of radioactive elements. However, some minerals often exhibit only small X-ray contrasts making them difficult to distinguish in 3D-scans. For example, quartz and plagioclase have closely similar X-ray characteristics, which limit the applicability of this approach. Here, we introduced a Monte Carlo simulation of beta particles in two dimensions, in order to calculate the dose rate from 2D images. The simulation results are compared between 2D modelling, 3D modelling and standard calculation. The implications for beta dose rate heterogeneity in grains are discussed. Using the “DosiVox-2D” software, we show that 2D scans may characterize complex heterogeneous 3D beta dose rate distributions. This approach could lead to a better understanding of micro-dosimetric phenomena and the improvements in the accuracy of dose rate calculation for heterogeneous samples. • We set up a new method of beta dose rate modelling for trapped charge dating. • A 2D representation of the sample is created from mineral mapping. • Beta dose rate is simulated in 2D and calculated for the different minerals. • 2D-modelling provides equivalent results to those of 3D-modellings. • 2D-modelling is easier to set up than 3D-modellings although less accurate.

Design-experimental assessment of radiological hazards in residential and industrial premises

Construction of residential and industrial buildings is associated with a wide application of both traditional and new structural and finishing materials. Radiological conditions in buildings are determined radiation of natural radionuclides contained in structural and finishing materials of buildings. Mostly around the world national criteria are used governing the content of natural radionuclides in building materials, based on the recommendations of international regulatory bodies. In Russia in order to justify admissibility of structural and finishing building materials in housing construction people use experimental methods. In European Union countries government suggests calculating estimation for assessment of the applicability different materials in various areas of housing and civil construction. Earlier, the authors established the fact of possible inconsistencies between Russian and European standards in different classes of materials. In order to compare the applicability of different approaches were held complex researches including experimental determination content of natural radionuclides in materials, calculation of dose rate in specific rooms and direct measurements of dose rates. As basic model authors used approach applied in the European Union countries. Experimental studies were performed using methods scintillation spectrometry and radiometry. The objects of research were premises with bearing and enclosing concrete structures faced with tiles. Experimental dose rate values were determined after deduction of the radiation background on the terrain. The conducted studies showed that the application of the computational model with the use of experimental values of the activity of natural radionuclides leads in some cases to incorrect values of dose rate in the premises, which is associated with the accepted value of the radiation background. Despite the simplicity of the proposed approach and its direct use can lead to unreliable results. So the coincidence of the calculated and experimental data is irregular. Obtained results allow to make the main conclusion about the limited application of the estimated approach to evaluation of radiological hazard in living quarters. The most reliable are the experimental approach, despite its great labor intensity. Calculate approach should be applied at the preliminary stage of rejection of materials.

Improvement of Sievert Integration Model in brachytherapy via inverse problems and Artificial Neural Networks

Increasing the radial distance, the accuracy of the Sievert Integration Model (SIM) decreases in a nonlinear manner, adding errors up of 10% into the dose rate calculations; a similar fact occurs to the 2D anisotropy function where the errors may achieve 30% as already was related. For that reason, this paper sought an innovative approach to optimize the error variance and its biases of dose rate calculations around a Nucletron brachytherapy source of 192Ir from 0 to 10 cm taken in the radial distance, using an improved SIM through a hybrid coupling of Artificial Neural Networks (ANNs) and Inverse Problem Theory (IPT). Since the traditional approach relies into the use of a small data set of dose rate, the ANNs generalized these doses, making possible to search more broadly optimum parameters to SIM using the IPT. The results showed excellent accuracy evaluated with the Root Mean Square Percentage Error (RMSPE). In conclusion, the low RMSPE values indicate that the methodology is consistent, showing an excellent agreement with the state of art of dosimetric measurement techniques.