FLUKA Monte Carlo Code Research Articles

Numerical estimation of calibration matrices for 241Am measurements using HPGe based in-vivo monitoring system

In-vivo measurements of 241Am using HPGe detectors become complicated when the active adjacent source organs interfere with the target organ measurements. It is important to calculate the contribution of confounding organs to estimate the activity of the target organ accurately. In the current study, numerical simulations were performed using FLUKA Monte Carlo code and International Commission on Radiological Protection (ICRP) computational reference phantoms to determine the calibration matrices consisting of the calibration and cross-talk coefficients for three interfering organs namely, lungs, liver and skeleton. It was found that the interference from adjacent organs contaminated with 241Am was found to be significant in the case of lungs and liver. Knee monitoring was least influenced by the activity possessed by other source organs due to their anatomical distance from the knee. A comparison between lung and liver coefficients obtained from the Lawrence Livermore National Laboratory (LLNL) torso phantom and thorax phantom derived from ICRP adult male voxel phantom was performed. It was found that variations in coefficients obtained from simulations and experiments range between 2% and 48%. The differences were attributed to the uncertainties arising from the composition of the phantoms and detectors, size and shape of organs, positional errors, and source distribution. A comparison of calibration matrices of adult male and adult female thorax voxel phantoms revealed that all the coefficients except knee as the target organ were larger for the female thorax phantom owing to the lesser bulk of attenuating tissues on its chest. The coefficients obtained from simulations for different phantoms also showed that the organ activity estimation can be significantly affected by the subject morphology.

Minimum dose path planning during reactor coolant system maintenance with corrosion product activity

To facilitate the periodic maintenance of the reactor coolant system, the nuclear power plant must be shut down. However, maintenance work on the primary loop components could result in post-shutdown radiation exposure. Activated corrosion products are the main source of radiation inside the containment building during post-shutdown. Traditionally, radiation detectors are used to measure in-containment radiation to protect workers from radiation risk during maintenance. However, these devices cannot predict corrosion product activity and the subsequent radiation dose. This study presents a minimum dose path planning methodology for protecting workers during the reactor coolant system maintenance in a typical pressurized water reactor nuclear power plant. Specifically, the study uses the corrosion product release and activation (CoPRA) code to estimate the corrosion product activity from the primary loop’s three major post-shutdown source terms. Subsequently, the FLUKA Monte Carlo code is used to simulate dose rates from the corrosion product activity, and the Three-Degree of Vertex shortest path algorithm is implemented to find the minimum dose path. Comparison results show that the predicted dose is consistent with the measured dose from the nuclear power plant. The results also show the flexibility and adaptability of the proposed method in finding the shortest path with the minimum dose in different reactor operating conditions.

Gamma photon and charged particle shielding features of praseodymium based glass matrix

Radiation shielding features of TePr glass system for helpful gamma-rays have been computed through FLUKA Monte Carlo code and Phy-X/PSD database. Linear attenuation coefficients (LAC) obtained with Phy-X/PSD and FLUKA code are found to be in good harmony with utmost relative difference < 3.7%. Based upon various parameters investigated, TePr1.2 (with highest amount of praseodymium) is found to be best shielding glass composition among the samples fabricated in this study for gamma-rays as well as charged particle beam of protons. The half value layer for the five glass samples between 0.284 and 2.51 MeV was estimated and we found that the HVL increases in the order TePr1.2, TePr1, TePr0.5, TePr0.2, TePr0. The proposed new glass system is found to have better shielding ability than commercial glasses and bismuth sodium borate glasses. The mass stopping power (MSP) for the tested glasses was reported, and we observed that MSP values for all the TePr glasses are not differing significantly. For protons, the SRIM based MSP values for TePr1.2 rise swiftly from 0.0612 to 0.135 MeV cm2/mg for kinetic energy range 0.01–0.14 MeV then decline slowly upto 0.0183 MeV cm2/ mg at 10 MeV.

Optical borophosphate glass system with excellent properties for radiation shielding applications

A new set of borophosphate glass samples: (50-x)P2O5–20B2O3–20CaO–10Na2O–xMoO3, where x = 0–15 mol. % were investigated for their radiation absorbing ability. The mass attenuation coefficient of the glasses were computed through XCOM calculations and photon transmission through the glasses via the FLUKA Monte Carlo code for gamma photons of energies 0.015 – 15 MeV. In addition, the stopping power and range of different charged particles were calculated from the ESTAR, PSTAR, ASTAR free online NIST database and SRIM code. Also, the fast (ΣR) and thermal neutron cross sections were calculated using standard expressions. We found that the photon attenuation ability of the PBCN-M glasses was improved with MoO3 content and the sample of PBCN-M4 is recorded as the best photon attenuator among them. The ΣR was within the range: 0.09285 – 0.09358 cm-1. The (ΣR)PBCN-M2 < (ΣR)PBCN-M4 < (ΣR)PBCN-M1 < (ΣR)PBCN-M3. The PBCN-M3 represented the optimum chemical structure of the glass system for fast neutron removal ability. In contrast to ΣR, σtot systematically increased as the glass density and MoO3 molar concentration increased. Comparing the photon shielding capability of the PBCN-M4 glasses with diverse shielding materials, suggested that investigated glass have huge potential to perform better. However, denser conventional shields showed superior ability to absorb photons compared to PBCN-M4 but the lower density and Pb-free nature of PBCN-M4 makes it more attractive for shielding applications.

Experimental and Simulated Spectral Gamma-Ray Response of a NaI(Tl) Scintillation Detector used in Airborne Gamma-Ray Spectrometry

Abstract. The objective of this work is to simulate the spectral gamma-ray response of NaI(Tl) scintillation detectors for airborne gamma-ray spectrometry (AGRS) using the state-of-the-art multi-purpose Monte Carlo code FLUKA. The study is based on a commercial airborne gamma-ray spectrometry detector system with four individual NaI(Tl) scintillation crystals and a total volume of 16.8 L. To validate the developed model, radiation measurements were conducted using 57Co, 60Co, 88Y, 109Cd, 133Ba, 137Cs and 152Eu calibration point sources with known activities and source-detector geometries under laboratory conditions. In addition, empirical calibration and resolution functions were derived from these measurements combined with additional radiation measurements adopting natural uranium, thorium and potassium volume sources. The simulation results show superior accuracy and precision compared to previous AGRS simulation models with a median relative spectral error &lt; 10 % for most of the radiation sources. Moreover, the implementation of a lower level discriminator model and detailed modelling of the laboratory result in a significant improvement in model accuracy for spectral energies &lt; 100 keV compared to previous studies. Yet thorough statistical analysis incorporating statistical and systematic uncertainty estimates revealed statistically significant deviations between the simulated and measured spectra in the spectral region around the Compton edge, which could be attributed to the scintillator non-proportionality. These findings imply that the linear energy deposition model applied in this and previously developed AGRS simulation models should be revised and considered to be replaced by more accurate non-proportional models.

Influence of combining Al2O3, La2O3, Gd2O3, and Dy2O3 with barium borosilicate glass-ceramics: a case study of gamma radiation interaction parameters

The goal of this investigation is to test the barium boro-silicate glass-ceramics with different additives against gamma radiation using the FLUKA Monte Carlo code. On four different glass-ceramics samples, the effect of an equal quantity of Al2O3, La2O3, Gd2O3, and Dy2O3 with barium borosilicate glass-ceramics on the radiation shielding capabilities of the glass-ceramics was investigated. In the examined glass ceramics, densities were obtained to investigate glass samples. The densities obtained were 3.92, 4.432, 4.52, and 4.88 g/cm3, respectively. BBSDy sample has the highest density which indicates that it is more effective for radiation shielding. The shielding parameters have been calculated at 0.356, 0.662, 1.173, and 1.333 MeV. The obtained results have been compared with the NistXCOM web page and Phy-X/PSD platform. The results showed a good agreement between FLUKA code, NistXCOM, and Phy-X/PSD. The calculated shielding parameters increase with additive (Al2O3, La2O3, Gd2O3, and Dy2O3). Moreover, the 50BaO–15SiO2–30B2O3–5Al2O3-5Dy2O3 specimen has the best radiation shielding features among the other glass-ceramics. In conclusion, the BBSDY sample containing 5 mol per cent 5 mol% Dy–III–Oxide would be the most effective in terms of radiation shielding, based on the results obtained in this study. When Dy-III Oxide concentrations were increased, linear and mass attenuation coefficient values were significantly increased, which contributed directly to the development of radiation shielding characteristics in the glass-ceramic.

The influence of beam optics asymmetric distribution on dose in scanning carbon-ion radiotherapy.

PurposeTo quantify the influence of beam optics asymmetric distribution on dose.MethodsNine reference cubic targets and corresponding plans with modulation widths (M) of 3, 6, and 9 cm and with center depths (CDs) of 6, 12, and 24 cm were generated by the treatment planning system (TPS). The Monte Carlo code FLUKA was used for simulating the dose distribution from the aforementioned original plans and the dose perturbation by varying ±5%, ±15%, ±20%, ±25%, and ±40% in spot full width half maximum to the X‐direction while keeping consistent in the Y‐direction. The dosimetric comparisons in dose deviation, γ‐index analysis, lateral penumbra, and flatness were evaluated.ResultsThe largest 3D absolute mean deviation was 15.0% ± 20.9% (mean ± standard deviation) in M3CD6, whereas with the variation from −15% to +20%, the values were below 5% for all cube plans. The lowest 2D γ‐index passing rate was 80.6% with criteria of 2%–2 mm by a +40% variation in M3CD6. For the M9CD24 with a −40% variation, the maximum 1D dose deviations were 5.6% and 15.7% in the high‐dose region and the edge of the radiation field, respectively. The maximum deviations of penumbra and flatness were 3.4 mm and 11.4%, respectively.ConclusionsThe scenario of beam optics asymmetric showed relatively slight influence on the global dose distribution but severely affected dose on the edge of the radiation field. For scanning carbon‐ion therapy facilities, beam spot lateral profile settings in TPS base data should be properly handled when beam optics asymmetry variation is over 15%.

Comparison of backscattered photons from various thicknesses of ordinary and baryte concretes using FLUKA code for energies 0.25–20 MeV

With the advancement of medical technology and increased radiation technologies, the main material in radiation building construction is concrete, either ordinary or, more recently, heavy-weight concrete called baryte concrete (BaSO4). Numerous studies were performed on calculating and measuring linear attenuation coefficients μ (cm−1) for ordinary and baryte concrete, but few for backscattered photons and reflection coefficients. As a result, determining the photon reflection coefficient and backscattered photon data for ordinary and baryte concrete has become more commonplace. The ability to provide precise reflection coefficient data has consequences in various fields, including radiation protection, industry, and radiotherapy room design. As a result, the purpose of this study is to determine photon reflection coefficients for different thicknesses of ordinary and baryte concrete. This new information might be applied to various uses, such as saving money and space by enhancing the design of mazes used in radiation facilities. The FLUKA Monte Carlo code was used to calculate photon reflection coefficients with different energies and for varying thicknesses of ordinary and baryte concrete to investigate the impact of concrete wall thicknesses. The reflection coefficient as a function of various energies has been tabulated and figured to study the dose reduction factor compared to ordinary concrete. Recent research reveals that baryte concrete with a high density is beneficial as a shield for constructing radiation facilities to minimise the backscattered photon exposure in the incident energy range below 2 MeV. This new data has been presented in many applications, such as improving the mazes used for radiotherapy rooms; reducing cost and space requirements. In addition, the Monte Carlo method enables calculating the energy distribution of reflected photons, and these were shown for a range of angles.

Mechanical and radiation shielding properties of CuO doped TeO2–B2O3 glass system

A series of borotellurite glass system doped with copper oxide was synthesized using the method of melt quenching. The density of the glass decreases whereas its molar volume increases in respect to copper oxide concentration. The amorphous nature of prepared glasses was proven with the utilisation of XRD spectra. Meanwhile, the glass samples' elastic properties are determined from the measurement of longitudinal and shear velocities using ultrasonic technique. The longitudinal modulus, shear modulus, bulk modulus and Young's modulus has shown an irregular trend that fluctuates within certain limits. Moreover, the FLUKA Monte Carlo code is implemented to determine the shielding features at gamma energy of 0.3565, 0.6616, 0.8348 and 1.3325 MeV. Linear attenuation coefficients acquired from FLUKA and XCOM are found to be in good agreement with relative difference <1%. The glass sample B2O3TeO2CuO1 (with highest percentage of TeO2) is found to be the best shielding glass material in this study.

Investigation and ANN-based prediction of the radiation shielding, structural and mechanical properties of the Hydroxyapatite (HAP) bio-composite as artificial bone

This paper focuses on the radiation shielding, structural and specifically mechanical features of the Hydroxyapatite (HAP) bio-composites which can be used instead of the bone and tooth tissues in the human body via FLUKA Monte Carlo Code (FMCC), Phy-X: PSD software, and an analytical method. Since HAP bio-composites are so brittle, their use is limited instead of bone in the human body. This challenging issue persuaded the scientists and researchers to solve the problem by inserting various oxide or dioxide materials into HAP bio-composites. Thus, in this work, TiO2 and CeO2 with different ratios of x = 0, 3, 7.5, 10 wt.% and y = 0, 2, 6, 7.5 wt.% are inserted into HAP bio-composites and thus eight types of S (S1, S2, S3, and S4) and B (B1, B2, B3, and B4) samples are produced. Using Artificial Neural Network (ANN), this study predicts and demonstrates the system's behavior. Outcomes reveal that increasing the TiO2 and CeO2 concentrations in the (100-x) HAP + xTiO2 and (100-y) HAP + yCeO2 bio-composites will improve the gamma photon shielding performance of the S and B samples. Furthermore, the photon and electron spatial maps for simulated geometries related to the S4 sample are extracted by the FLUKA Monte Carlo Code and are represented graphically. The produced electrons with the highest energy are monitored in lead volumes due to various interactions of gamma photons with lead shields. In addition, sharp peaks are reported for Zeff curves related to the B samples which may be due to the K-edge absorption of the Ce in HAP samples. The FLUKA results are in full agreement with the predicted targets via the ANN algorithm. Moreover, increasing the CeO2/TiO2 concentrations in HAP bio-composites will enhance the rigidity of the chosen S and B samples. The rising percentage of the mechanical moduli related to the S and B series vary between 30% and 90% which may be due to the relationship between the density of the selected HAP bio-composites and mechanical moduli.

MRI-LINAC dosimetry approach by Monte Carlo codes coupling charged particle radiation transport with strong magnetic fields

This study investigates the dosimetric effects due to the strong magnetic fields in the MRI guided radiotherapy. FLUKA and PENELOPE Monte Carlo codes were used to study the deflection of the electron trajectories, while evaluating dosimetric effects, considering both simplified arrangements and complex realistic clinical applications setups. Simulation codes proved to be suitable, and comparisons with analytic approaches were satisfactory, thus remarking non-negligible dosimetry effects as a consequence of the presence of strong magnetic fields.

Radiation Protection at the Large Hadron Collider: Problematics, Challenges and Advanced Monte Carlo Simulation Techniques

This paper provides an overview of the problems, challenges, and the advanced simulation techniques used to study and plan complex interventions in radiation areas at CERNs Large Hadron Collider and its future upgrade to the High-Luminosity Large Hadron Collider. The operational radiation protection aspects are supported by state of the art simulations by means of the FLUKA Monte Carlo code and estimates conducted via other tools such as ActiWiz and SESAME, used within the HSE-RP group.

Study on the measurement of high energy neutrons using Zirconium capped CR-39 (Zr-CR-39) combined detector - Monte Carlo simulation and experimental validation

Neutron measurement, be it dose or fluence is a challenging field in high energy accelerators where the neutron energy varies from thermal to high-energy neutrons usually extending beyond 20 MeV. CR-39 SSNTD (Solid State Nuclear Track Detector) used in neutron measurements has near flat energy response in the range of 500 keV to 10 MeV. By capping the CR-39 with a suitable converter such as Zirconium (Zr), it's response can be enhanced in the higher neutron energy ranges (beyond 10 MeV). Zr has significant (n,2n) cross-section of ∼1 b in the neutron energy range of 10–20 MeV. Hence, when high energy neutrons impinge on Zr, neutrons of lower energy are produced by (n,2n) reactions, which create additional recoil proton tracks in CR-39. At higher neutron energies (>20 MeV) additional tracks can also be created due to other charged secondaries such as deuteron, triton, alphas and recoil carbon and oxygen. The enhancement in track density of CR-39 in presence of Zr as compared to that of bare CR-39 is termed as the Total Enhancement Factor (TEF). Monte Carlo simulations with FLUKA and GEANT4 codes are used to optimize the Zr converter thickness with 14 MeV and 20 MeV, and are also verified with measurements. At an optimum converter thickness of 1 mm, FLUKA and GEANT4 Monte Carlo codes were used to observe the enhancement in the response of the combined detector (Zr-CR-39) against the bare detector for different mono-energetic neutron beams in the range 14 MeV - 1 GeV. The enhancement obtained for Zr converter are also compared with Lead (Pb) converter. Experimental studies were carried out to validate the simulations by irradiating Zr-CR-39 to 14 MeV, 20 MeV and high energy neutrons spectra available at CERF facility. Measured TEF values are 1.27, 2.5 and 7.85 for 14 MeV, 20 MeV and CERF neutrons spectra respectively. Results of simulations and measurements are found to be in agreement within 20%.

Optical properties and radiation shielding competence of Bi/Te-BGe glass system containing B2O3 and GeO2

In this study we present the optical features including several parameters, gamma-ray photon and fast neutron absorption parameters of quaternary glasses with the chemical structure: xBi2O3 + (80-x)TeO2 + 10B2O3 + 10GeO2 for x = 40, 50, and 60 mol% and coded as Bi/Te-BGe1, Bi/Te-BGe2, and Bi/Te-BGe3 respectively. The FLUKA Monte Carlo and XCOM code were used to evaluate the photon shielding parameters of the glasses for energies between 0.284 and 2.506 MeV. The optical quantities such as the molar refractivity Rm, molar polarizability αmol, reflectance loss (Rloss), optical transmission (Topt), metallization criterion (M) and the static dielectric constant εstaof the glasses were estimated to understand their optical characteristic. Mass attenuation coefficient of 0.345, 0.368, and 0.386 cm2/g was obtained for Bi/Te-BGe1–3 respectively at 0.284 MeV. Evaluated Zeff ranged from 30.94 to 53.22, 32.73 – 56.29, 34.52 – 58.75 for the Bi/Te-BGe1–3 glasses respectively. The HVL within the investigated energy similar to other estimated photon shielding parameters showed great enhancement as the density and Bi2O3 content of the glasses improved. Similar improvement on the fast neutron removal cross section also increased from 0.1141 to 0.1178 cm−1 as bismuth oxide increases in the glass system. The estimated optical parameters also showed great dependence on the chemical structure of the glasses. The present Bi/Te-BGe1–3 glasses showed excellent fast neutron and photon absorption capacity compared to some common shields; this makes them potentially useful for radiation protection applications.

The response of various neutron dose meters considering the application at a high energy particle accelerator

Abstract The applicability of several neutron detectors for dose measurements at a neutron field typical for high energy particle accelerators is investigated. The response of four commercially available active neutron dose meters and two passive detectors to neutrons from a 241Am-Be(α,n) source and to neutrons at the CERN EU high energy reference field was determined experimentally and simulated using the Monte-Carlo code FLUKA. Fluence response functions and dose responses for the different detectors were calculated in the energy range between 1 keV and 10 GeV. The results show that the dose response to the high energy neutron field at CERN of the conventional rem-counters is lower by a factor of 2 to 2.5 if compared to the dose response to a 241Am-Be(α,n) neutron source. The rem-counters exhibiting an additional layer of lead inside the moderating structure showed dose readings which differ only up to 25 %. A thermoluminescent based neutron detector was tested for comparison. These passive detectors revealed a neutron response similar to a rem-counter and may be preferable for situations with highly pulsed beams.

Optical and radiation shielding studies on tellurite glass system containing ZnO and Na2O

Three TeO2-ZnO-Na2O ternary glass samples with chemical formula 80TeO2-xZnO-(20-x)Na2O (where x = 20, 10, and 0 mol% represents TeZnNa1, TeZnNa2, and TeZnNa3 respectively) were investigated for their optical and radiation shielding attributes. The calculated optical parameters namely structure molar refractivity (RMol), molar polarizability (αMol.), reflection loss (RLoss) and optical transmission (TOpt.) all vary with the chemical content of the glasses. The FLUKA Monte Carlo code and XCOM software was used to estimate mass attenuation coefficient MAC of the glasses. Furthermore, the fast neutron absorption probability ΣR was estimated through standard expressions. The MAC values varies between 0.03796 and 0.16191 for TeZnNa1, 0.03793–0.16206 for TeZnNa2 and 0.03791–0.16221 cm2/g for TeZnNa3 as photon energy changes from 284 keV to 2.506 MeV. Analysis of other photon shielding parameters such as effective atomic number and mean free path showed that TeZnNa3 is the best shield among the glasses. The ΣR of the glasses also changed from 0.0942 to 0.1117 cm−1 as Na2O content increased from 0 to 20 ml%. The photon and fast neutron removal cross section of the glasses compared relatively well with standard radiation shields. The glasses are hence, can be used for radiation absorption purposes.

Actinides induced irradiation damage and swelling effect in irradiated Zircaloy-4 after 30 years of storage

Abstract. After several years in the reactor core, irradiated nuclear fuel is handled and subsequently stored for a few years under water next to the core, to achieve thermal cooling and decay of very short-lived radionuclides. Thereafter, it might be sent to dry-cask interim storage before final disposal in a deep geological repository. Here, the spent nuclear fuel (SNF) is subject to a series of physicochemical phenomena which are of concern for the integrity of the nuclear fuel cladding. After moving the SNF from wet to dry storage, the temperature increases, then slowly decreases, leading the hydrogen in solid solution in the cladding to precipitate radially with consequent hydride growth and cladding embrittlement (Kim, 2020). Another phenomenon affecting the physical properties of the cladding during interim dry storage is the irradiation damage produced in the inner surface of the cladding by the alpha decay of the actinides present at the periphery of the pellet, particularly when the burnup at discharge is high. SNF pellets with high average burnup present larger fuel volumes at the end of their useful life due to accumulation of insoluble solid fission products and noble gases, which leads to disappearance of the as-fabricated pellet–clad gap. Further swelling is expected as a consequence of actinide decay and the accumulation of helium. This leads to larger cladding hoop stress and larger alpha decay damage. The present work first investigates the variation in diameter caused by pellet swelling in an irradiated Zircaloy-4 cladding after chemical digestion of the uranium oxide (UOx) pellet. Second, the irradiation damage produced during the 30 years elapsed since the end of irradiation in terms of displacements per atom (dpa) is studied by means of the FLUKA Monte Carlo code. The irradiation damage produced by the decay of actinides in the inner surface of the cladding extends for less than 3 % in depth. The considered cladded UOx pellet was extracted from a pressurized water reactor (PWR) fuel rod consisting of five segments, with an average burnup at discharge of 50.4 GWd (tHM)−1.

Neutron spectra at a high energy heavy ion accelerator measured with a TLD-based Bonner spectrometer.

A Bonner sphere spectrometer with TLD cards as sensitive elements was constructed and used at the GSI accelerator complex to measure the neutron spectral fluence rates for radiation safety purposes. The neutron spectra were measured in the synchrotron tunnel during 550 MeV/u 40Ar beam operation and in Cave A experimental area with a 1 GeV/u 56Fe beam impinging on a thick aluminum target. Unfolded spectra were compared with those calculated using the Monte Carlo code FLUKA. The measured spectra show a good agreement with the simulations and the set proved itself to be a useful tool for different radiation safety purposes.

Dynamic Response of Graphitic Targets with Tantalum Cores Impacted by Pulsed 440-GeV Proton Beams

Various graphite targets with a tantalum core were exposed to 440 GeV pulsed proton beams at the HiRadMat facility at CERN. The dynamic response was investigated by monitoring the surface velocity of the samples by laser Doppler vibrometry. The study comprises different graphite grades, such as polycrystalline, expanded and carbon-fiber reinforced graphite, and low-density graphitic foams, all candidates for beam-intercepting devices in high-power accelerators. The purpose of the tantalum core is to concentrate the large energy deposition in this high-density material that withstands the localized beam-induced temperature spike. The generated pressure waves are estimated to result in stresses of several hundred MPa which subsequently couple with the surrounding graphite materials where they are damped. Spatial energy deposition profiles were obtained by the Monte Carlo code FLUKA and the dynamic response was modelled using the implicit code ANSYS. Using advanced post-processing techniques, such as fast Fourier transformation and continuous wavelet transformation, different pressure wave components are identified and their contribution to the overall dynamic response of a two-body target and their failure mode are discussed. We show that selected low-intensity beam impacts can be simulated using straight-forward transient coupled thermal/structural implicit simulations. Carbon-fiber reinforced graphites exhibit large (macroscopic) mechanical strength, while their low-strength graphite matrix is identified as a potential source of failure. The dynamic response of low-density graphitic foams is surprisingly favourable, indicating promising properties for the application as high-power beam dump material.

Monte Carlo simulation of photons backscattering from various thicknesses of lead layered over concrete for energies 0.25\u201320\xa0MeV using FLUKA code

There is an increased interest in determining the photon reflection coefficient for layered systems consisting of lead (Pb) and concrete. The generation of accurate reflection coefficient data has implications for many fields, especially radiation protection, industry, and radiotherapy room design. Therefore, this study aims to calculate the reflection coefficients of photons for various lead thicknesses covering the concrete. This new data for lead, layered over concrete, supports various applications, such as an improved design of the mazes used for radiotherapy rooms, which helps to reduce cost and space requirements. The FLUKA Monte Carlo code was used to calculate photon reflection coefficients for a concrete wall with different energies. The reflection coefficient was also calculated for a concrete wall covered by varying thicknesses of lead to study the effect of lining this metal on the concrete wall. The concrete's reflection coefficient data were compared to internationally published data and showed that Monte Carlo calculations differed significantly from some of the extrapolated data. The absorbed dose of backscattered photons for various thicknesses of lead covering the ordinary concrete has been tabulated as a function of the reflection angle. Also, the reflection coefficient as a function of the Pb thicknesses covering the ordinary concrete has been figured to study the dose reduction factor. The generation of accurate data for reflection coefficients is vital for many fields, especially for radiation protection and radiotherapy room design. The new data have been presented for lead layered over concrete in various applications, such as an improvement in the design of the mazes used for radiotherapy rooms, thereby reducing the cost and space requirements. In addition, the Monte Carlo method enables calculating the energy distribution of reflected photons, and these were shown for a range of angles.