Geant4 Code Research Articles

Mass attenuation coefficients, effective atomic numbers and electron densities of some contrast agents for computed tomography

In the clinical computed tomography (CT) examinations, a photon is attenuated as it passes a patient by tissues and contrast agents (CAs). The CAs can increase the visibility of internal structures or fluids within the patient. In this work, we have investigated the photon interaction parameters of some CT contrast agents such as iotrolan, iodixanol, iohexol, ioxilan, ioversol, and iomeprol. The mass attenuation coefficients (μ/ρ) of these contrast agents have been determined using Geant4 code in the energy range from 1 keV to 1 MeV for total photon interaction. The validity of the Geant4 code was verified by comparing the simulation results with those calculated by the XCOM program. A very good agreement was observed between μ/ρ values obtained by both Geant4 and XCOM codes. The μ/ρ values were then used to estimate the effective atomic numbers (Zeff) and electron densities (Neff) for the selected CT contrast agents. It was found that the values of μ/ρ, Zeff and Neff depend on the photon energy and increase with increasing iodine concentration in the composition of CAs. Also, the Zeff values were observed in the range of 6–50 and the Neff values were observed in the range of 2 – 21 (1023 electron/g). The present study would be helpful to develop new CT contrast agents to serve in vivo imaging applications.

New Hybrid 3D Analytical Linear Energy Transfer (LET) Calculation Algorithm Based on the Pre-Calculated Data from Monte Carlo (MC) Simulations

The traditional 1D analytical LET model of the spot scanned proton therapy (SSPT) assumes a uniform LET distribution in the lateral direction for each depth, which will under-estimate the LET value in the low dose penumbra region and over-estimate the LET value in the high dose region compared with Monte Carlo (MC) simulations. We propose a fast and accurate hybrid 3D analytical LET calculation method based on the pre-calculated data from MC simulations. The method is analogous to the pencil beam algorithm to calculate the physical dose in proton therapy with the 3D LET calculation kernel generated by MC as follows: Step-1: Using a well-benchmarked MC code (Geant4) to generate LET distributions of single energy proton beams in water for all 97 energies and derive the dose-averaged LET (LETd) lateral profiles at various depths starting from the surface with 1mm depth interval increase until the depth with the 1% of the peak dose after the Bragg peak. Step-2: Using a customized “error function” to fit the aforementioned LET lateral profiles for every depth of every energy, and store the fitted coefficients as a lookup table. Step-3: During the dose/LET calculation, the stored fitted coefficients and the fitting function will be used to calculate the LETd values based on the spot energies and the water equivalence thickness (WET). The inhomogeneity is handled by taking WET into account in both beam and lateral directions. We then validated the improvement of our new method by comparing the calculated LET distributions with the results calculated from the MC and the previous 1D analytical LET model in twelve patients with different disease sites. Compared with the MC simulation results, the LETd calculation of our new method is systematically better than the traditional 1D model, with the average 3D gamma analysis results of all 12 patients with the threshold of 3% and 2mm increased from 94% to 98% in the region with dose >10% of the maixumum dose. The majority improvement comes from the more accurate LETd calculation in the low dose penumbra region (5%-20% of the prescription dose). In some cases, our new method only has less than 3% relatively difference compared with the MC simulation in the penumbra region, while the previous 1D model under-estimates the LETd value by about 20% to 50%. The calculation time of our new method is comparable to our previous 1D analytical LET calculation algorithm. Our new hybrid 3D analytical method (1) can calculate LETd in SSPT accurately and efficiently, (2) will be used for biological effects evaluation and LET-guided robust optimization in SSPT to improve the OAR protection, (3) is easily to be integrated by other analytical dose engine.

Estimates of TPR spectrometer instrumental signal-to-background ratios and count rate limits for ITER like plasmas

The work presented is a realistic simulation of the response function for a detection efficiency optimized Thin-foil proton recoil (TPR) neutron spectrometer. The TPR spectrometer consists of a thin foil acting as neutron-to-proton converter followed by ΔE-E detectors operating in coincidence mode. In this work, two different spectrometer designs were considered using segmented silicon detectors. Design 1 has slightly better resolution while design 2 is more compact and has higher efficiency. The TPR spectrometer response functions were simulated in the energy range 8–18 MeV in steps of 40 keV for the two designs using the dedicated Monte Carlo code GEANT4. The resulting simulated response functions were broadened using experimentally determined energy resolutions of the detectors, in order to produce more realistic response functions. Using these broadened response functions together with an ITER like neutron spectrum and neutron induced background simulations ΔE/E energy deposition plots were created. The energy-cuts, for 14 MeV neutron signal identification, were applied to the ΔE-E plots leading to an estimate of the expected signal-to-background ratio. In addition, pile-up fraction and maximum expected count rates were estimated. Results show that the ΔE-E energy cuts show a great prospect of increasing the signal-to background ratio for the TPR spectrometer. In addition the TPR spectrometer has energy resolution (FWHM/E) of around 5% for 14 MeV neutrons for both investigated designs. The spectrometer can cope with maximum count rate expected and have a sufficient signal-to-background ratio in the neutron energy range of interest to perform fuel ion ratio measurements. However an increase of acquisition channels would be beneficial to limit the pile-up rate.

X-ray photons attenuation characteristics for two tellurite based glass systems at dental diagnostic energies

X-ray photons attenuation characteristics for the two tellarite based glasses Bi2O3– B2O3– TeO2– TiO2 and PbO–ZnO–TeO2–B2O3 have been investigated at dental diagnostic energies (between 30-80 keV) using Geant4 code and WinXcom software. The correlation coefficient (R2) is utilized to evaluate the extent to which Geant4 results are related to the WinXcom data. For the both series, R2 is close to 1 for all samples and this implies a perfect degree of association between the Geant4 and WinXcom data. The linear attenuation coefficient is proportionally increased with addition of TeO2 in both series, which implies that there is a decreasing tendency in the X-ray photon transmission corresponding with an increase in the TeO2 content in the glasses. The half value layer (HVL) decreases as the density increases and this decreasing is very notable at 70 and 80 keV. The maximum HVL for all samples occurs at 80 keV and this implies that the HVL gradually increases as the energy of the X-ray photons increase. Also, the increment of TO2 in the glasses (in both systems) leads to reduce the mean free path and BiTeTi6 and PbTeB6 samples have the lowest MFP. The MFP for both systems was compared with three heavy concretes and the comparison revealed that the selected systems can be utilized to fabricate protection masks used during diagnostic radiation of the head or oral cavity.

The influence of PbO on the radiation attenuation features of tellurite glass

Tellurite based glasses have been of considerable interest in the last decades due to their interesting mechanical, physical and chemical features. Presently, different investigators and engineers are being devoted for developing the Tellurite based glasses to be used as effective radiation shielding materials. Our aim is to study the energy as well as PbO content dependence on the gamma rays attenuation behavior of lead tellurite glasses when the PbO content changes from 0 to 30 mol% between 0.356 and 2.506 MeV photon energies. This was achieved by considering the lead tellurite glass system with the composition x(PbO)-(100-x)(TeO2). The photon attenuation factors of the lead tellurite glasses were evaluated at five energies utilizing Monte Carlo simulation (Geant4 code). The reative difference values between Geant4 and XCOM data for the six glasses are less than 7% and this validates the accuracy of the obtained results. The mass attenuation coefficient (μ/ρ) reduces with an increase in the energy whereas it increases when TeO2 is replaced by PbO. The influence of PbO on the μ/ρ is significant only at 0.356 and 0.511 MeV, but at the other energies it is found that PbO slightly affects the attenuation factor. The linear attenuation coefficient of 30PbO–70TeO2 (coded as TePb6) is higher than the other samples. The effective atomic number has the maximum value at 0.356 MeV (in the range of 27.76–41.25). We compared the mean free path for the tellurite glasses with RS-520 and RS-360, Osmani, nasir and PHP glasses. The results reveal that the lead tellurite glasses (especially the samples with high PbO content) are a candidate as protection materials from hazardous radiation and have better attenuation in comparison to some nuclear engineering materials.

An investigation on physical, structural and gamma ray shielding features of Dy3+ ions doped Telluroborate glasses

Structural, physical and gamma ray shielding properties of Dy3+ doped telluroborate glasses have been discussed. The molar volume of the glasses is proportional to the TeO2 content. The FTIR studies investigate the expansion of TeO2 content and the dynamic change of tetrahedral [BO4] units into trigonal [BO3] units. Using the absorption spectra, optical band gap (Eg) and Urbach's energy (ΔE) values are calculated. The molar volume of oxygen (VO) and boron‑boron separation (<dB–B>) values are increasing while there is a fall in oxygen packing density (OPD) from 0% to 40% of TeO2. The ionic nature of the present glasses is analyzed from the bonding parameter (δ), optical basicity (Λth), covalent and ionic characteristic factors. The gamma photon attenuation investigated employing GEANT4 code. The transmission factor (TF) shows that 0TBD sample (0 wt% of TeO2) has the highest TF, while the 4TBD sample (40 wt% of TeO2) has the lowest TF.

Comparison between MCNP5, Geant4 and experimental data for gamma rays attenuation of PbO–BaO–B2O3 glasses

Monte Carlo simulations, MCNP5 and Geant4 codes were developed to investigate radiation shielding properties of xPbO–(50-x) BaO–50B2O3 (where 5 ≤ x ≤ 45 mol%) consider to be glass systems. The mass attenuation coefficients were evaluated for different PbO concentration in the glass samples for varies photon energies of 0.356, 0.662, 1.173 and 1.332 MeV. The obtained mass attenuation coefficient values used to calculate half-value layer, effective atomic number, and electron density. The simulation parameters were compared with experimental data. Results show that the simulation results of mass attenuation coefficients for all PbO concentrations were generally in good agreement with experimental results, however, mass attenuation coefficient values calculated using Geant4 were slightly lower than MCNP5 and experimental data on the low energy of 0.356 MeV. The results obtained from MCNP5 and Geant4 codes might be able to assessment mass attenuation for different glass systems. Furthermore, gamma ray, fast neutron and charged particle interaction for the glass systems were studied using buildup factors, fast neutron removal cross sections and ranges respectively.

Comprehensive study on the structural, optical, physical and gamma photon shielding features of B2O3-Bi2O3-PbO-TiO2 glasses using WinXCOM and Geant4 code

Glass samples of titanium-lead bismuth borate with the composition of (60B2O3 – 20Bi2O3 – (20-x)PbO – xTiO2), where 0 ≤ x ≤ 10 wt% labeled as BBPT0.0 (x = 0) to BBPT10.0 (x = 10) were synthesized. The melt quenching technique was used in the preparation. The amorphous character and vibrational groups for all samples were confirmed through XRD and FTIR measurements, respectively. The optical band gap (Eopt), Urbach's energy (Eurbach), and the optical dispersion constants have been determined using UV–Vis spectra. Results reveal that the values of the optical band gap energy exhibit a decrease with the TiO2 concentration, in contrast with Urbach's energy which displays a linear increase from 0.43 to 0.64 eV. In addition, the values of single-oscillator energy (Eo) has the same behavior of the optical energy gap (Eopt), while the dispersion energy parameter (Ed) has an inversely behavior. Mass attenuation coefficients (μ/ρ) of the investigated glasses have been computed using the Geant4 simulation toolkit and WinXCOM database software. Geant4 model of a HPGe detector has been employed to perform the Monte Carlo simulations for the glass samples with five thicknesses from 1 to 3 cm at the gamma ray energies in the range of 356–2510 keV. The transmission fractions as a function of the thickness at 356 keV photon energy for all studied glasses has been performed. Half value layer (HVL) was derived from the μ/ρ values. Results showed the studied glasses have good shielding capability according to their mass attenuation coefficients, transmission fractions and HVL values.

Synthesis, physical, structural and shielding properties of newly developed B2O3–ZnO–PbO–Fe2O3 glasses using Geant4 code and WinXCOM program

New zinc lead borate glasses with (50B2O3–10ZnO–(40 − y) PbO)–yFe2O3: 0 ≤ y ≤ 10 wt%) composition were synthesized by the conventional melt-quench technique. The amorphous nature of the glasses was tested by using X-ray diffraction spectroscopy. FTIR spectra for the glasses were recorded in the range of 4000–400 cm−1. Physical properties of the proposed glasses were evaluated. The mass attenuation coefficients (μ/ρ) for the glasses were calculated by the Geant4 simulation toolkit code and XCOM software within the photon energy range 365–2510 keV. Different γ-ray shielding parameters such as half-value layer, tenth value layer, and transmission fraction for the proposed glasses were evaluated using the obtained values of the (μ/ρ). Results reveal that for 1 cm glass thickness, the (50B2O3–10ZnO–30PbO)–10Fe2O3 sample has the highest transmission, due to the lower amount of Pb, and thus poorer shielding capability among the other glass samples. In addition, the probabilities of γ-ray interaction with the studied glasses were investigated.

The C6D6 detector system on the Back-n beam line of CSNS

The neutron capture cross sections are very important in the field of nuclear device design and basic physics research. Hydrogen-free liquid scintillator such as C6D6 detectors are widely used in the neutron capture cross-sectional measurements for the low neutron sensitivity and fast time response. The Back-n white neutron source at China Spallation Neutron Source is the first spallation white neutron source in China, and it is suitable for neutron capture cross-sectional measurement. A C6D6 detector system was built in the Back-n experimental station. The pulse height weighting technique was used to determine the system’s detection efficiency. The response to gamma rays of the C6D6 detector was measured, and the energy resolution function was determined. Monte Carlo simulation with Geant4 code was carried out to get the weighting function of this C6D6 detector system. Additionally, the systematic uncertainty of the weighting function was also determined. According to the experimental and simulation results, this C6D6 detector system can be used to measure neutron capture cross section.

Investigation ofthe gamma ray shielding parameters of (100-x)[0.5Li2O–0.1B2O3–0.4P2O5]-xTeO2 glasses using Geant4 and FLUKA codes

In the present work, the radiation shielding proeprties of lithium borophosphtellurite glass system with the composition of (100-x) [0.5Li2O-0.1B2O3–0.4P2O5]-xTeO2 (x = 0, 10, 20, 30 and 40 mol%) were reported. The mass attenuation coefficient (μ/ρ) of the selected glass system (LiBPTe0, LiBPTe10, LiBPTe20, LiBPTe30 and LiBPTe40) was calculated at 356 keV, 662 keV, 1173 keV and 1330 keV photon energies by using Geant4 and FLUKA codes and the results were compared to the values obtained theoretically using WinXcom software. The correlation theory has been used to calculate the correlation coefficients (R2) between the results from WinXcom and the both simulations codes (Geant4 and FLUKA) for each sample. Moreover, some other radiation shielding parameters such as effective atomic number, photon transmission, mean free path and half value layer were calculated. The gamma photon transmission results showed that the transmission values of the 356 keV photons are lower than those of energies 662, 1173 and 1330 keV. At 356 keV, the transmission value for LiBPTe0 (at 6 cm thickness) is found to be 0.245 while at the same energy and thickness the transmission value is 0.088 for LiBPTe40. Out of all the glasses considered in present work, LiBPTe40 possesses the highest values of effective atomic number (Zeff). The Zeff values for the present glasses are 8.02, 9.55, 11.16, 12.85 and 14.64 for LiBPTe0, LiBPTe10, LiBPTe20, LiBPTe30 and LiBPTe40 respectively at 356 keV. The mass attenuation coefficient and effetive atomic number results showed that utilization 40% mol of TeO2 significantly improves the gamma ray shielding properties of the selected glass system. The shielding effectiveness of the selected samples was compared with other reference materials namely two radiation shielding glasses and four concretes used in different shielding purposes in terms of the mean free path (MFP) at 356 keV. The results revealed that the MFP of RS-253-G18 is lower than LiBPTe0sample. Moreover, the MFP of ordinary concrete is higher than the MFP of all glasses under examination, thus the investigated tellurite glasses have better attenuation features than the ordinary concrete.

Olivine mineral used in concrete for gamma-ray shielding

The aim of this study is assessing the concretes containing olivine, which is an abundant and locally mined mineral in Turkey, in terms of several essential photon shielding parameters, which are total mass attenuation coefficients (μ/ρ), mean free paths (mfp), half and tenth value layer thicknesses (HVL and TVL, respectively), effective atomic numbers (Zeff), and electron densities (Nel). For this purpose, two types of concretes with and without olivine content have been produced and studied by photon transmission experiments, simulations and theoretical calculations. In the experimental part, photon transmission technique has been performed using a 16k 3″ × 3″ NaI(Tl) detector system and so, μ/ρ, mfp, HVL, and TVL values of concretes have been determined experimentally at 511 and 1275 keV photon energies. Aforesaid parameters have also been estimated by simulations at the same energies using Geant4 Monte Carlo code. For theoretical calculations, the web version of NIST XCOM photon cross-section library has been utilized. By this way, all parameters have been calculated at 1 keV–100 GeV energy region and compared with data from experiments and simulations. It has been observed that olivine addition affected the aforementioned radiation shielding parameters of concretes positively at 511 and 1275 keV energies. Also, results from each method have been found to be in a reasonably good agreement, especially for the ones from XCOM and Geant4. So, it has been concluded that Geant4 code is a powerful tool in such studies.

Determination of 18F positron range in PET imaging using Monte Carlo simulation - Geant4 code

Objective: To determine the distribution of the positron range of radionuclide 18F using Monte Carlo Simulation. Method: 18F is widely available for routine clinical use. In this work, we perform a theoretical calculation of the distribution of the 18F positron range using Monte Carlo (MC) simulation. To collect a statistically significant sample of distance values between the beta emitter point and the Line Of Response (LOR), approximately50000 tracks in water were generated and propagated until annihilation. The radial cumulative probability distribution G2 D(δ) as a function of was adjusted and the analytical formulae was obtained. Finding: The maximum, mean and 1D values of the positron range for 18F simulated and propagated in water are calculated and compared with other studies. The application of Monte Carlo simulation for positron range calculation emphasizes the adoption of this calculation for the radionuclide properties and their propagated medium. The calculation results of beta particle energy loss are in good agreement with many studies that dealt with the same matter having the same energy range. The bin size, source shape, simulation code and propagation medium are the main parameters responsible on the slight differences in the positron range calculation results. Application: It is well known from most existing studies that PET resolution blurring is coupled to the positron range and that it is important to take this range into account in the image reconstruction process. In this study, we calculated the positron range of 18F in water, which is not an obvious affect and it is related to the values of Emean and Emax of 18F 0.250 MeV and 0.635MeV, respectively. These values will cause positron particles to propagate in the medium with a range not exceeding 2 mm with an FWHM(px) is 0.16 mm and FWTM(px) is 1.05 mm. Keywords: Geant4 Code, Positron Emission Tomography, Positron Range

A novel methodology to assess linear energy transfer and relative biological effectiveness in proton therapy using pairs of differently doped thermoluminescent detectors

A new methodology for assessing linear energy transfer (LET) and relative biological effectiveness (RBE) in proton therapy beams using thermoluminescent detectors is presented. The method is based on the different LET response of two different lithium fluoride thermoluminescent detectors (LiF:Mg,Ti and LiF:Mg,Cu,P) for measuring charged particles. The relative efficiency of the two detector types was predicted using the recently developed Microdosimetric d(z) Model in combination with the Monte Carlo code PHITS. Afterwards, the calculated ratio of the expected response of the two detector types was correlated with the fluence- and dose- mean values of the unrestricted proton LET. Using the obtained proton dose mean LET as input, the RBE was assessed using a phenomenological biophysical model of cell survival.The aforementioned methodology was benchmarked by exposing the detectors at different depths within the spread out Bragg peak (SOBP) of a clinical proton beam at iThemba LABS. The assessed LET values were found to be in good agreement with the results of radiation transport computer simulations performed using the Monte Carlo code GEANT4. Furthermore, the estimated RBE values were compared with the RBE values experimentally determined by performing colony survival measurements with Chinese Hamster Ovary (CHO) cells during the same experimental run. A very good agreement was found between the results of the proposed methodology and the results of the in vitro study.

Escape-suppression shield detector for the MINIBALL $ \gamma$γ-ray spectrometer

A bismuth-germanate (BGO) escape-suppression shield for the high-purity germanium triple-cluster detector of the MINIBALL \( \gamma\)-ray spectrometer was designed and built. Monte Carlo simulations with the simulation code GEANT4 were performed to guide the construction and to determine the detector geometry of the new BGO shield. After the first measurements concerning mechanical properties of the BGO housing and the performance of the photomultiplier tubes at the Institut de Physique Nucleaire, Orsay, the prototype BGO escape-suppression shield was combined with a MINIBALL triple-cluster detector at the Institut fur Kernphysik, Cologne. A dedicated electronics and digital data-acquisition system was put into operation in order to determine timing properties of the combined coincidence measurement and to measure values for the energy resolution of the BGO detectors, for the BGO low-energy threshold, and for the crucial peak-to-total ratio (P/T). The measured P/T value for a standard 60Co \( \gamma\)-ray source compares well with expectations and will allow to proceed with the amendment of the MINIBALL triple-cluster detectors with an escape-suppression shield for improved in-beam \( \gamma\)-ray spectroscopy especially at the new HIE-ISOLDE accelerator for radioactive ion beams at CERN.

Numerical Simulation and Sensitivity Study of the Rhodium Self-Powered Neutron Detector Used in PWR

Focusing on the specific self-powered neutron detector (SPND) using rhodium as the emitter material, this paper performs the numerical simulation and sensitivity study for SPNDs used in a typical pressurized water reactor (PWR). The computational model was developed based on a PWR lattice code bamboo lattice and a Monte Carlo code Geant4. In order to validate the model, the neutron sensitivities with single-energy incident neutrons are calculated and compared against the experimental results. Influences of the incident neutron energy spectrum and the emitter burnup to the neutron sensitivity of the SPND are analyzed. Bamboo lattice is used to calculate the incident neutron energy spectrum and the emitter nucleus densities which are the environment of the SPND. Numerical results show that the neutron sensitivity decreases due to the increase of the enrichment, the assembly average burnup, the boron concentration, the moderator temperature, and the burnable poison rods. The increase in the emitter burnup will lead to the decrease of the neutron sensitivity. Taking the incident neutron energy spectrum and the nucleus density into account together, the neutron sensitivity decreases with the increase in the assembly average burnup.

МОДЕЛИРОВАНИЕ ИСТОЧНИКА НЕЙТРОНОВ НА ОСНОВЕ УСКОРИТЕЛЕЙ ЭЛЕКТРОНОВ В GEANT4

Monte Carlo calculations with geant4 code provides a convenient tool for studying the parameters of neutron sources. The purpose of this work was simulation of the 30 to 140 MeV electron beam transport through a tungsten target with the geant4 program code. In addition, the neutrons yield produced in the target from the electron beam at the IREN JINR facility was studied. The geant4 program code correctly describes the neutron photoproduction process during the primary electron beam interactions with a tungsten target. To verify the cross sections in geant4 at high photon energies, the cross sections in the TALYS program code were calculated. Used cross sections are in good agreement with previous experimental data. The obtained photoneutron yield dependences from the electron energy for a target simulating an IREN facility are from 7.0·10-3 at 30 MeV to 4.2·10-2 neutrons per electron at 200 MeV.

Structural investigation of RBMK nuclear graphite modified by 12C+ ion implantation and thermal treatment

Graphite is a major core component in the graphite-moderated RBMK-1500 nuclear reactor. Neutron irradiation induces radiation damage in the graphite matrix, which affects physical properties of material. In order to generate defects that could be similar to those induced by neutrons in the RBMK-1500 nuclear reactor, the raw RBMK graphite samples were implanted with 12C+ ions at the energy of 700 keV at varying fluences ranging from 1.2 × 1015 to 1.2 × 1016 ions/cm2. The ion beam interaction simulations were performed by using SRIM-2013 and GEANT4 codes. In order to study temperature effects on the evolution of the ion implanted graphite structure the thermal treatment was applied. The structural disorder level on the surface of RBMK graphite samples was investigated by using Raman spectroscopy before and after ion implantation as well as after thermal treatment. The decrease of the graphitic sp2-related content as well as formation of an amorphous structure were observed on the surface of the ion implanted samples, while thermal treatment resulted in significant reordering at the ion implantation fluence of 7.2 × 1015 ions/cm2 and lower. The understanding of structural dynamics of nuclear graphite under irradiation conditions provides insights into migration of impurities (e.g. 14N) as well as distribution of activation products (e.g. 14C, 36Cl) in the irradiated graphite matrix.

Applications of Gamma-ray Transport Modeling to Improving Borehole Logging for Uranium

Borehole logging for uranium was much studied in the 1970s and early 1980s. Calibration required building large concrete calibration pits. Codes for gamma-ray transport modeling were then the specialty of large nuclear research institutions. Today, building concrete pits is difficult but sophisticated modeling codes are freely available. For example, the GEANT code, used in this study and available from CERN, enables gamma-ray problems with complex geometry to be evaluated by modeling rather than experiment. This paper illustrates three examples of recent applications of the modeling approach to examine deconvolution, the effect of high U grades on total count logging tools and the applicability of the cerium-doped, lanthanum bromide (BrilLanCe) detectors to direct U logging using the 1.00 MeV peak from 234mPa.

New high temperature resistant heavy concretes for fast neutron and gamma radiation shielding

Abstract In the present work, we developed three new high temperature resistant heavy concretes as novel radiation shielding materials. For this purpose, chrome ore (FeCr2O4), hematite (Fe2O3), titanium oxide (TiO2), aluminum oxide (Al2O3), limonite [FeO (OH) nH2O], siderite (FeCO3), barite (BaSO4), nickel oxide (NiO) materials and alumina cement were used. Mass combination ratios of components and total macroscopic cross sections (scattering, absorption, capture, fission) of the samples were calculated by using GEANT4 code. The resistances of the prepared samples were evaluated in terms of compression strength after exposure at the 1000 °C temperature. Neutron equivalent dose rate measurements were carried out by using 4.5 MeV 241Am-Be neutron source and BF3 detector. All results were compared with normal weight concrete and paraffin. The results of neutron dose indicate that neutron absorption ability of the new heavy concretes is higher than normal weight concrete and paraffin. In addition to neutron measurements, different γ-ray shielding parameters such as mass attenuation coefficient (MAC), effective atom numbers (Zeff), half value layer (HVL) and mean free path (MFP) have been calculated using WinXCOM software in order to investigate the effectiveness of using the prepared concretes as a radiation shielding materials. Gamma-ray results were compared with concretes and Pb-based glass.