Borated Polyethylene Research Articles

Dose equivalent consideration from neutron contamination in modified radiotherapy vault: a Monte Carlo study

Laminated barriers incorporating metal sheets provide effective protection for space-restricted radiotherapy centers. This study aimed to assess photoneutron contamination in smaller vaults protected by different compositions of multilayer barriers during simulated pelvic radiotherapy with 18 MV photon beams. Monte Carlo Simulations of 18 MV LINAC (Varian 2100 C/D) and Medical Internal Radiation Dose (MIRD) phantom were used to assess photoneutron contamination within reconstructed vaults incorporating different combinations of metal sheet and borated polyethylene (BPE) during pelvic radiotherapy. The findings highlight a 3.27 and 2.91 times increase in ambient neutron dose Hn* (10) along the maze of reconstructed vaults that use lead and steel sheets, respectively, compared to concrete. The Hn* (10) outside the treatment room increased after incorporating a metal sheet, but it remained within the permissible limit of 20 μSv/week for uncontrolled areas adjacent to the LINAC bunker, even with a workload of 1000Gy/week. Neutron equivalent doses in the patient’s organs ranged from 0.22 to 0.96 mSv Gy−1. There is no notable distinction in the organ’s neutron equivalent dose, fatal cancer risk, secondary radiation-induced cancer risk, and cancer mortality for various laminated barrier compositions. Furthermore, the use of metal sheets for vault wall reconstruction keeps the variation in cancer risk induced by photoneutrons below 6%, while risks of fatal cancer and cancer mortality vary less than 11%. While the metal portion of the laminated barrier raises the neutron dose, the addition of a BPE plate reduces concerns of increased effective dose and secondary malignancy risk.

Highly Efficient and Eco-Friendly Thermal-Neutron-Shielding Materials Based on Recycled High-Density Polyethylene and Gadolinium Oxide Composites.

Due to the increasing demands for improved radiation safety and the growing concerns regarding the excessive use of plastics, this work aimed to develop effective and eco-friendly thermal-neutron-shielding materials based on recycled high-density polyethylene (r-HDPE) composites containing varying surface-treated gadolinium oxide (Gd2O3) contents (0, 5, 10, 15, and 20 wt%). The results indicate that the overall thermal-neutron-shielding properties of the r-HDPE composites were enhanced with the addition of Gd2O3, as evidenced by large reductions in I/I0, HVL, and TVL, as well as the substantial increases in ∑t and ∑t/ρ of the composites. Furthermore, the results reveal that the values for tensile properties initially increased up to 5-15 wt% of Gd2O3 and then gradually decreased at higher contents. In addition, the results show that the addition of Gd2O3 particles generally increased the density (ρ), the remaining ash at 600 °C, and the degree of crystallinity (%XC) of the composites. This work also determined the effects of gamma irradiation on relevant properties of the composites. The findings indicate that following gamma aging, the tensile modulus slightly increased, while the tensile strength, elongation at break, and hardness (Shore D) showed no significant (p < 0.05) differences, except for the sample containing 5 wt% of Gd2O3, which exhibited a noticeable reduction in elongation at break. Furthermore, by comparing the neutron-shielding and mechanical properties of the developed r-HDPE composites with common borated polyethylene (PE) containing 5 wt% and 15 wt% of boron, the results clearly indicate the superior shielding and tensile properties in the r-HDPE composites, implying the great potential of r-HDPE composites to replace virgin plastics as effective and more eco-friendly shielding materials.

Design and implementation of the monochromator shielding for the cold neutron spectrometers XINGZHI and BOYA

An innovative monochromator shielding is designed and implemented for the cold neutron spectrometers XINGZHI and BOYA operated by Renmin University of China at China Advanced Research Reactor. Via Monte Carlo simulations and careful mechanical designs, a shielding configuration has been successfully developed to satisfy safety requirements of below 3 μSv/h dose rate at its exterior, meanwhile fulfilling space, floor load and nonmagnetic requirements. Composite materials are utilized to form the sandwich-type shielding walls: the inner layer of boron carbide rubber, the middle layer of steel-encased lead and the outer layer of borated polyethylene. Special-shaped liftable shielding blocks are incorporated to facilitate a continuous adjustment of the neutron energy while preventing radiation leakage. Our work has demonstrated that by utilizing composite shielding materials, along with the sandwich structure and liftable shielding blocks, a compact and lightweight shielding solution can be achieved. This enables the realization of advanced neutron scattering instruments that provide expanded space of measurement, larger energy and momentum coverage, and higher flux on the sample. This shielding represents the first of its kind in neutron scattering instruments in China. Following its successful operation, it has been subsequently employed by other neutron instruments across the country.

Neutron transmission analysis in borated polyethylene, boron carbide, and polyethylene: Insights from MCNP6 simulations

This study delves into the neutron transmission properties of three pivotal materials: borated polyethylene, boron carbide (B4C), and polyethylene, employing the advanced MCNP6 (version 6.2) Monte Carlo code. The primary objective was to discern the impact of boron content and material density on neutron transmission efficacy. Results highlighted the formidable neutron shielding prowess of B4C, which exhibited the lowest neutron transmission factor among the studied materials. Borated polyethylene samples revealed an inverse correlation between boron concentration and neutron transmission, emphasizing boron's paramountcy in neutron attenuation. In contrast, polyethylene showcased notable neutron moderating capabilities, necessitating synergistic integration with materials possessing higher neutron absorption cross-sections for superior neutron shielding. Material density and the presence of other elemental constituents further influenced neutron transmission. The research provides a comprehensive understanding of neutron-material interactions, setting the groundwork for enhanced safety measures in nuclear applications.

Simulation research on the characteristics of a minitype neutron reference radiation field

In order to research the characteristics of minitype neutron reference radiation field regulated by ISO 8529 series standards, Monte Carlo simulation method was applied to calculate the ratio of room-scattered neutron response to the direct neutron response, neutron fluence and energy spectra in the radiation field. Firstly, the characteristics of the field were investigated from the four physical factors as shielding materials, field sizes, shielding thicknesses and borated polyethylene with different boron contents. Secondly, the response of sphere instrument with different diameters in the field and the influence of instrument on the field were analyzed. The results showed that minitype neutron reference radiation field with a size of 2 m × 2 m × 2 m, a shielding material of 5 % borated polyethylene and a shielding thickness of 10 cm was expected to be used for the calibration of neutron measuring instrument.

Design of beam line for BNCT applications in HEC-1 channel of IRT-T research reactor

The feasibility of using the HEC-1 beam port of the IRT-T research reactor to generate an epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) was investigated using MCNPX2.6 Monte Carlo simulation. The reactor was first simulated to design and optimize a Beam Shaping Assembly (BSA) that meets the neutron beam criteria recommended by the International Atomic Energy Agency (IAEA). The suggested BSA configuration consisted of a cylindrical geometry with 25 cm of MgF2 and 15 cm of Fluental® as a moderator, 15 cm of Pb as a reflector, two 5 cm Bi slabs as gamma-ray shields, and 4 mm of 6Li and 5 mm of borated polyethylene sheets as thermal neutron filters. The results showed that the epithermal neutron flux at the BSA exit was 2.3×109 n/cm2.s, and in-phantom dose analysis indicated that the designed beam could be used for the treatment of deep brain tumors within an allowable treatment time of 50 min.

Characterization of the Radiation Contamination in an 18MV Linac Treatment Room Made of some Nanoparticles Mixture

Purpose: Shielding against radiation in radiotherapy and radiology requires deep knowledge of radiation physics and shielding design methods. The application of nanoparticles in the photon and neutron dose moderation has been proven in the literature.&#x0D; Materials and Methods: Effective neutron mass removal cross-section (ΣR/ρ) was for the ordinary concrete doped with 50nm in diameter nanosphers mixture of Fe2O3(5%), WO3(5%), B4H (5%), Pb2O3(5%) was estimated with MCNP5 Monte Carlo (MC) code and N-XCOM computational program. An 18MV linac room made of the nanoparticles dopped ordinarily simulated. Additionally, the room was considered with three legs in the maze and photoneutron and capture γ-ray Dose Equivalent (DE) were estimated at the modeled rooms maze.&#x0D; Results: Total ΣR/ρ with energies 100 keV-2000keV was estimated using MC and N-XCOM as 0.02802-0.02687 cm2/gand 0.02810- 0.02687 cm2/g, respectively. Total ΣR/ρ of the neutron for the pure ordinary concrete was estimated by MC code and N-XCOM with energies 100 keV-2000keV as 0.02802-0.02687 cm2/g and 0.02810- 0.02687 cm2/g. Borated Polyethylene (BPE) and Lead required thickness calculated as 7.43×10-06 mm and 4.73×10-06 mm for the capture γ-ray shielding.

Feasibility study of nanomaterials synthesis at MNSR research reactor through design and construction of a gamma irradiation cell

Herein, a gamma irradiation cell consisting of a borated polyethylene (BPE) cylinder with a cadmium inner wall was designed and constructed at Isfahan miniature neutron source reactor (MNSR) for synthesis of metal nanoparticles. The cell is placed in a dry channel irradiation site near the reactor core. The performance of this irradiation cell with appropriate shape and size was evaluated and the neutron flux (φ) and the neutron and gamma dose rates (Dn and Dg) inside it were calculated using MCNP code. The results showed that the designed cell could deliver an appropriate gamma dose rate to samples and reduce the thermal neutron flux from 5.82 × 1010 to 6.28 × 107n.cm−2.s−1, i.e. over 99.89% attenuation. Also, the gamma dose rate inside the designed neutron shield is calculated to be 3.83 kSv.h−1. To evaluate the designed and constructed cell, synthesis of Pd and Ag nanoparticles (NPs) on SiO2 supporter was done. Prepared nanocomposites were characterized by using XRD, FE-SEM, and EDXA techniques. Observed results showed metallic Pd and Ag NPs were produced successfully employing a designed gamma irradiation cell at the reactor via gamma-assisted route in nanometer sizes for the first time.

Gamma ray diagnostics for high time resolution measurement in large helical device

A 1-inch LaBr3:Ce gamma ray scintillation detector, characterized by high time and energy resolution, was installed to advance energetic particle physics studies in Large Helical Device (LHD). We reduced the size of the scintillator from 3 inches to 1 inch to reinforce the radiation shielding to reduce the unwanted signal in the detector induced by fast neutrons and stray gamma rays according to the commissioning results. The radiation shielding composed of 10% borated polyethylene and lead was redesigned to suppress the gamma-ray induced signal based on the Monte Carlo three-dimensional neutron and gamma-ray transport code MCNP6. We increased the lead thickness from 50 mm to 77.5 mm to suppress the stray gamma ray effect. The gamma ray spectrum was measured in the hydrogen neutral beam heated deuterium plasma with 6LiF pellet injection. We might obtain a gamma ray peak near 0.48 MeV due to the 6Li(d,p'γ)7Li reaction.

Overview of nuclear data production system at RAON

Nuclear Data Production System (NDPS), a fast neutron facility for nuclear science and applications, was constructed at the Rare Isotope Accelerator complex for ON-line experiments (RAON) in Korea. NDPS is designed to provide both white and quasi-monoenergetic neutrons using 98 MeV deuteron and 20 – 83 MeV proton beams with thick graphite and thin lithium targets, respectively. Neutron energy is determined by employing the Time-Of-Flight (TOF) technique, along with a pulsed deuteron (or proton) beam with a repetition rate of less than 200 kHz. Fast neutrons are produced in the target room and are guided to the TOF room through a 4 m long neutron collimator consisting of iron and 5% borated polyethylene. The neutron beam is monitored using a parallel plate avalanche counter (PPAC) and a micro-mesh gaseous (MICROMEGAS) detector installed in the TOF room, so as to measure the energy and the position of neutrons.

Design of a radiation shield for an HPGe detector for a reactor-based PGNAA facility

The purpose of this study was to design an appropriate neutron and gamma radiation shield for the HPGe detector to be used in the PGNAA facility in the Isfahan MNSR reactor. In order to improve the parameters related to shield design, including the thickness of the material used in the design, the different type of materials considered and the relative reduction of neutron and gamma dose, shields were simulated by MCNPX code and the results of calculations were compared. It was concluded that the optimal thickness of lead as gamma absorber is 4 cm and the optimal thickness of borated polyethylene as neutron absorber is 10 cm. In this case, the neutron and gamma dose rate at the location of the detector inside the shield was equal to 5.11 μSv.h−1 and 7.5 μSv.h−1, respectively. Also, according to the investigations conducted for different materials as neutron absorbers, it was found that the best absorber for neutrons is borated polyethylene, which absorbs more neutrons and produces less secondary gamma than other materials and the best absorber for gamma is bismuth. The shielding has brought the relative percentage of neutron and gamma dose reduction about 90%.

COMBINED METHOD OF BULK MATERIAL SHIELDING EVALUATION FOR 200 MEV HIGH ENERGY NEUTRON SOURCE USING PHITS MODELLING AND PARTIAL DENSITY

Neutron encounters difficulties in shielding protection. Thus, many researchers have performed simulation and experimental research on neutron shielding materials. The characteristic of materials is highly dependent on neutron energy. The evaluation of neutron shielding for various materials, such as iron, concrete, aluminum, and borated polyethylene (BPE), was conducted in this paper through simulation using a Monte Carlo code of PHITS 3.27 and calculation via partial density method. A mono-energetic neutron source with an energy of 200 MeV is emitted perpendicular to the shielding material with a thickness of 105 cm. The parameters measured in this analysis include flux, fast neutron removal cross-section, neutron depth dose, ambient dose H*(10) equivalent, and neutron dose reduction factor (RF). Results show that iron is a good material against high-energy neutron and secondary photon radiation at the energy range with the highest removal cross-section and the lowest RF value (0.39), followed by concrete, BPE, and aluminum. The integrated fluence and effective dose profiles were consistent with previous results in the literature. Benchmarking calculation of neutron dose RF was conducted with other publications and was in good agreement within the value range.

Radionuclide activity in cryoconite from glaciers of the Central Caucasus, Russia

This work presents the results of a study of radionuclide activity in cryoconite from glacier and glacial soil of the Central Caucasus. Cryoconite were sampled from the surface of the Garabashi glacier, and soil samples were taken from the humus horizon of periglacial mountain forest-meadow soil. Measurements were performed with low-background germanium gamma spectrometers located inside a passive shield consisting of ∼20 cm of copper, ∼15 cm of lead, and ∼8 cm of borated polyethylene. The specific activity of radionuclides (Be-7, K-40, Th-232, U-235, U-238, Cs-137) was established. It was revealed that all measured spectra contain γ-lines from decays of K-40, decay chains of U-238, U-235 and Th-232. In addition, the spectra of cryoconite samples from the Garabashi glacier show a 477.6 keV line from the decay of the cosmogenic isotope Be-7, and in the soil sample a 661.7 keV line from the radionuclide Cs-137. No radionuclide Be-7 was detected in the mountain forest-meadow soil. Radionuclide Cs-137 is present in the soil sample, while in cryoconite, is not detected. Radioisotope activities in the Garabashi glacier cryoconite, except for the cosmogenic isotope Be-7, do not differ significantly in terms of mass, i.e., the content of K-40, U-238, U-235 and Th-232 in them is approximately the same. The activity of all the studied radionuclides in the soil sample compared to cryoconite samples is lower, although the differences are not significant, except for Th-232, whose activity in soil is almost two times lower.

Development of a low-background neutron detector array

A low-background neutron detector array was developed to measure the cross section of the $$^{13}$$ C( $$\alpha$$ ,n) $$^{16}$$ O reaction, which is the neutron source for the s-process in AGB stars, in the Gamow window ( $$E_\text {c.m.}$$ = 190 ± 40 keV) at the China Jinping Underground Laboratory (CJPL). The detector array consists of 24 $$^{3}$$ He proportional counters embedded in a polyethylene cube. Owing to the deep underground location and a borated polyethylene shield around the detector array, a low background of 4.5(2)/h was achieved. The $$^{51}$$ V(p, n) $$^{51}$$ Cr reaction was used to determine the neutron detection efficiency of the array for neutrons with energies $$E_\text {n}<$$ 1 MeV. Geant4 simulations are shown to effectively reproduce the experimental results. They were used to extrapolate the detection efficiency to higher energies for neutrons emitted in the $$^{13}$$ C( $$\alpha$$ ,n) $$^{16}$$ O reaction. The theoretical angular distributions of the $$^{13}$$ C( $$\alpha$$ ,n) $$^{16}$$ O reaction were shown to be important in the estimation of the uncertainties of the detection efficiency.

Research on the shielding performance and optimization of new type foam metal matrix composite shielding materials

The research and development of new composite shielding materials is of great practical significance for radiation shielding of reactors. Foam metal is light in weight, and has the characteristics of high specific strength and high specific area. From the structural characteristics of the foam metal, a new type of foam metal matrix composite shielding material can be prepared. The foam metal matrix composite shielding material consists of two parts: metal matrix and filling material. In this paper, there are three kinds of foam metal matrix (RAFM, tungsten and nickel) and eight kinds of filling materials (B4C, three kinds of lead-boron polyethylene compounds, two kinds of borated polyethylene compounds, polyimide resin, air). The shielding effect of the above-mentioned foam metal matrix composite shielding material on neutrons and gamma at different porosities is explored through MCNP simulation, and several high-performing shielding combination schemes are obtained. The neutron transmittance less than 50% and the gamma ray transmittance less than 30% are used as shielding targets. The shielding combination schemes obtained above are applied to the shielding model, and then the minimum thickness of the shielding target can be achieved by simulation. The results show that when the foam metal matrix is tungsten, the filling material is B4C and the porosity is 80%, its thickness of 8.5CM can reach the shielding target, which is the best shielding scheme. Compared with the traditional iron/water mixed shielding, the thickness reduction rate of this solution is 84.45% and the mass reduction rate is 47.53% when the above-mentioned shielding target is reached. In addition, the above optimal scheme is compared with the experimental data of nickel-based foam metal to further illustrate the reliability of the new foam metal matrix composite shielding material.

Investigation of Dose Rate Distribution in an Experimental Hall of a RIKEN Accelerator-Driven Compact Neutron Source Based on the ⁹Be(p, n) Reaction With 7 MeV Proton Injection

To characterize the dose rate distribution in an experimental hall of a RIKEN accelerator-driven compact neutron source (RANS) based on the ⁹Be(p, n) reaction with 7 MeV proton injection, systematical measurements and calculations for neutron and gamma-ray dose rates by GEometry ANd Tracking (GEANT), Particle and Heavy Ion Transport code System (PHITS), and Monte Carlo N-Particle (MCNP) codes were performed. Calculations always underestimated measurements when proton beam loss effect was not considered. Relatively good agreements were observed among the different simulation codes. To explain the underestimations, the additional dominant neutron and gamma-ray sources due to proton beam loss were identified at the position around exit of the drift tube linac (DTL), made of copper, and the beam pipe from quadrupole (Q) magnets to steering (ST) magnets, made of aluminum, from measurements with placing collimators along linac. The beam loss fractions of 2&#x0025;&#x2013;3&#x0025; on copper and 1&#x0025; on aluminum, respectively, were the most appropriate estimation. In addition, we proposed the possible measures to reduce the measured total dose rate of <inline-formula> <tex-math notation="LaTeX">$3.8~\mu $ </tex-math></inline-formula>Sv/h at the operator position in the control room, with the addition of a wall at the entrance of experimental hall and extension of borated polyethylene (BPE) at the end of the beam. As a result, the dose rate became 2.5 times lower than the current one.

Design of a neutron shielding performance test system base on Am–Be neutron source

A system for neutron shielding performance test was designed based on an Am–Be neutron source and 3He proportional counter. The Am–Be neutron source was shielded by paraffin wax block with a thickness of dozens of centimeters and collimated by polyethylene rod. The neutron energy spectra in front of the collimator were simulated by MCNP. The detector system was designed based on 3He proportional counter and different response tuning assembly (RTA) to optimize the neutron energy region of interest. A method for testing neutron shielding performance for thermal, epithermal, fast neutrons and neutron ambient dose equivalent rate was proposed by employing different response tuning assembly (RTA). Common radiation shielding materials including polyethylene (PE), borated polyethylene (BPE) and graphite were conducted with the designed system. The results demonstrated that the proposed method is efficient for investigating the neutron shielding performance of materials.

Tailoring neutron beam properties by target-moderator-reflector optimisation

Compact accelerator-driven neutron sources allow to operate multiple optimised target-moderator-reflector (TMR) units adapted to the requirements of the respective instruments. The compact design of the TMR units allows an efficient coupling of neutron production, neutron moderation and extraction, but requires a novel way of optimisation. The neutronic performance of different TMR units based on polyethylene, heavy water and a mixture of heavy and light water moderators together with Pb and Be reflectors and a borated polyethylene absorber is discussed. Extraction channels for thermal and cold neutrons are investigated regarding the energy and time spectra.

Hybrid halide perovskite neutron detectors

Interest in fast and easy detection of high-energy radiation (x-, γ-rays and neutrons) is closely related to numerous practical applications ranging from biomedicine and industry to homeland security issues. In this regard, crystals of hybrid halide perovskite have proven to be excellent detectors of x- and γ-rays, offering exceptionally high sensitivities in parallel to the ease of design and handling. Here, we demonstrate that by assembling a methylammonium lead tri-bromide perovskite single crystal (CH3NH3PbBr3 SC) with a Gadolinium (Gd) foil, one can very efficiently detect a flux of thermal neutrons. The neutrons absorbed by the Gd foil turn into γ-rays, which photo-generate charge carriers in the CH3NH3PbBr3 SC. The induced photo-carriers contribute to the electric current, which can easily be measured, providing information on the radiation intensity of thermal neutrons. The dependence on the beam size, bias voltage and the converting distance is investigated. To ensure stable and efficient charge extraction, the perovskite SCs were equipped with carbon electrodes. Furthermore, other types of conversion layers were also tested, including borated polyethylene sheets as well as Gd grains and Gd2O3 pellets directly engulfed into the SCs. Monte Carlo N-Particle (MCNP) radiation transport code calculations quantitatively confirmed the detection mechanism herein proposed.

Optimization of the electron beam dump for a GeV-class laser electron accelerator

The advances of laser-driven electron acceleration offer the promise of great reductions in the size of high-energy electron accelerator facilities. Accordingly, it is desirable to design compact radiation shielding for such facilities. A key component of radiation shielding is the high-energy electron beam dump. In an effort to optimize the electron beam dump design, different material combinations have been simulated with the FLUKA Monte Carlo code in the range of 1–40 GeV. The studied beam dump configurations consist of alternating layers of high-Z material (lead or iron) and low-Z material (high-density concrete or borated polyethylene) in either three-layer or five-layer structures. The designs of various beam dump configuration have been compared and it has been found that the iron and concrete stacking in a three-layer structure with a thick iron layer results in the lowest dose at 1, 10, and 40 GeV. The performance of the beam dump exhibits a strong dependence on the selected materials, the stacking method, the beam dump thickness, as well as the electron energy. This parametric study provides general insights that can be used for compact shielding design of future electron accelerator facilities.