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This study aimed to fabricate six different Copper(II) oxide reinforced Calcium-Borate glasses with different types of substitutions such as Al2O3/V2O5, BaO2/V2O5, and ZnO/V2O5. Accordingly, a depth characterization process has been performed for ACV, BCV and ZCV glasses with a nominal composition of 55B2O3–35CaO–9Al2O3-0.5CuO-0.5V2O5, 55B2O3–35CaO-8.5Al2O3-0.5CuO–1V2O5, 55B2O3–35CaO–9BaO2-0.5CuO-0.5V2O5, 55B2O3–35CaO-8.5BaO2-0.5CuO–1V2O5, 55B2O3–35CaO–9ZnO-0.5CuO-0.5V2O5, 55B2O3–35CaO-8.5ZnO-0.5CuO–1V2O5. Optical, structural, physical, and experimental neutron/gamma shielding properties of synthesized glasses were determined, respectively. Experiments measuring neutron exposure indicated how well glass samples attenuated fast neutrons. The RADACS software was used to record data from a BF3 gas proportional detector from the Canberra NP-100B series and a 241Am/Be neutron source with a 10 mCi activity. The absorption edge belonging to the samples are found between 420 nm and 480 nm. Our findings showed that the optical band gaps for the samples ranged from 1.179 to 2.022 eV. The wavenumber range of 400–1600 cm-1 was evaluated for the resulting peaks. The region with the highest band formation was approximately 760–1170 cm-1. While BCV0.5 and BCV1 glasses with 9% and 8.5% BaO2 insertion have the largest MAC values ranging between 2.255–0.076 and 2.156–0.076 cm2/g, the lowest MAC values varying between 0.361–0.0761 and 0.366–0.076 cm2/g belong to ACV0.5 and ACV1 glasses with 9% and 8.5% Al2O3 addition. Our results showed that the BCV glass family has superior material properties among the fabricated glasses. It can be concluded that BaO2/V2O5 glasses may be used in replacement of Copper(II) oxide glasses to provide monotonic behavior on crucial characteristics while maintaining the greatest density increase for large gamma ray shielding properties.

Long turn-off time limits high frequency operation of Bipolar Junction Transistors (BJTs). Turn-off time decreases with increases in the recombination rate of minority carriers at switching transients. Fast neutron irradiation on a Si BJT incurs lattice damages owing to the displacement of silicon atoms. The lattice damages increase the recombination rate of injected holes with electrons, and decrease the hole lifetime in the base region of pnp Si BJT. Fast neutrons generated from a beryllium target with 30 MeV protons by an MC-50 cyclotron were irradiated onto pnp Si BJTs in experiment. The experimental results show that the turn-off time, including the storage time and fall time, decreases with increases in fast neutron fluence. Additionally, it is confirmed that the base current increases, and the collector current and base-to-collector current amplification ratio decrease due to fast neutron irradiation.

This study discusses the effect of using 232Th instead of 238U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO2 with (Th, 235U)O2 and (Th, 233U) O2. Firstly, the infinity multiplication factor (Kinf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd2O3 and Er2O3 on the Kinf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of 232Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of 135Xe, 149Sm, and 151Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρCR), the effective delayed neutron fraction βeff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

PURPOSE A new type of electronic dosimeter is presented, capable of discerning between the doses of gamma photons and neutrons in a mixed beam as found in Boron Neutron Capture Therapy (BNCT). We introduce a real-time dosimeter based on a thick gate Field Oxide Field Effect Transistor (FOXFET) covered with a neutron converter layer containing gadolinium. METHODS To sensitize the FOXFET dosimeter to neutron fluxes, a converter layer containing gadolinium oxide particles embedded in photoresines was deposited over the sensor surface. Mixed neutron-gamma field configurations with different neutron energy spectra were used to assess the FOXFET response, considering different thicknesses of the neutron converter layer. RESULTS The total gamma sensitivity of the devices resulted to be 43 mV/Gy. The responses of sensors with different converter layer thicknesses irradiated with different neutron spectra are simulated using GEANT4 code. The response to photons is not significantly modified with thin conversion layers when used in water medium. CONCLUSIONS A real-time dosimeter comprising a pair of FOXFET sensors-only one of them with a gadolinium neutron converter layer -allows the simultaneous measurement of gamma dose and neutron flux during BNCT irradiations. This article is protected by copyright. All rights reserved.

Journal of radioanalytical and nuclear chemistry (2021). doi:10.1007/s10967-021-08102-2, Published by Springer Science + Business Media B.V., Dordrecht [u.a.]

10B-neutron capture was observed optically using a boron-added liquid scintillator. Trimethyl borate was dissolved in a commercially available liquid scintillator at natural boron concentrations of approximately 1 wt% and 0.25 wt%. The boron-added liquid scintillator was placed in a phantom quartz bottle and irradiated by thermal neutrons (~ 105 n/[cm2 s]) for 150, 300, and 600 s. The luminescence of the liquid scintillator was clearly observed using a cooled charge-coupled device (CCD) camera during irradiation. The luminance value recorded by the CCD camera was proportional to the duration of irradiation by thermal neutrons. The luminescence distribution showed reasonable agreement with that of energy deposition by Li and alpha particles from 10B-neutron capture reactions calculated via Monte Carlo simulations. When trimethyl borate was not dissolved in the liquid scintillator (0 wt% natural boron), no visible luminescence was observed even after 600 s of irradiation. These findings demonstrate that the observed luminance originates from the Li and alpha particles generated by 10B-neutron capture reactions. Consequently, the luminescence distribution is directly related to the boron dose of the liquid scintillator. To the best of our knowledge, direct experimental optical observations of boron dose distribution have not yet been reported. This novel technique will be useful for quality assurance in boron neutron capture therapy (BNCT) because instantaneous neutron irradiation may be sufficient for the observing the intense neutron beam used in clinical BNCT (~ 109 n/[cm2 s]), and quick evaluation of the boron dose distribution is expected to be feasible.

Boron neutron capture therapy (BNCT) was first proposed as early as 1936, and research on BNCT has progressed relatively slowly but steadily. BNCT is a potentially useful tool for cancer treatment that selectively damages cancer cells while sparing normal tissue. BNCT is based on the nuclear reaction that occurs when 10B capture low-energy thermal neutrons to yield high-linear energy transfer (LET) α particles and recoiling 7Li nuclei. A large number of 10B atoms have to be localized within the tumor cells for BNCT to be effective, and an adequate number of thermal neutrons need to be absorbed by the 10B atoms to generate lethal 10B (n, α)7Li reactions. Effective boron neutron capture therapy cannot be achieved without appropriate boron carriers. Improvement in boron delivery and the development of the best dosing paradigms for both boronophenylalanine (BPA) and sodium borocaptate (BSH) are of major importance, yet these still have not been optimized. Here, we present a review of this treatment modality from the perspectives of radiation oncology, biology, and physics. This manuscript provides a brief introduction of the mechanism of cancer-cell-selective killing by BNCT, radiobiological factors, and progress in the development of boron carriers and neutron sources as well as the results of clinical study.

The TALYS 1.9 program is used to calculate cross sections of processes that occur during the scattering of fast neutrons on 48Ti, 52Cr, and 56Fe isotopes. Results from model calculations are compared to data obtained in the TANGRA project using tagged neutrons from the yields of γ-quanta for Ti, Cr, and Fe isotopes, and the results from earlier experiments. Analysis of the model description’s sensitivity to the choice of the direct reaction mechanism shows the most informative characteristic from this viewpoint is the differential cross section of inelastic scattering.

Neutron response functions of a silicon sensor, which is applied to a new real-time personal albedo neutron dosemeter, have been simulated for low energy neutrons from 0.01 eV to 10 keV using the Monte Carlo technique. The angular neutron response functions were obtained by multiplying simulated neutron energy spectra crossing the neutron sensor and the cross-section of the 6Li(n,t)4He reaction. The neutron response functions have been closed to the dose conversion coefficient of personal doses, Hp(10) recommended by the International Commission on Radiological Protection by selecting neutrons incident from angles of 105° to 180° with respect to an axis perpendicular to an acrylic phantom surface. From these simulation results, the neutron energy response function has been improved by shielding the sensor with cadmium box without a face toward the phantom. The neutron sensor provides a good conformance to the Hp(10) conversion coefficients within 15% for low energy neutrons.

In this paper, the performance of SiC-based neutron detector was investigated using GEANT4 toolkit. Thin-film-coated and trench type neutron detectors were considered and their intrinsic detection efficiency were calculated according to the relevant structural parameters, with Low-Level Discriminator (LLD) value fixed at 300 keV. For thin-film-coated detector, the analysis carried out for varied layer thicknesses of 10B and 6LiF showed that the intrinsic detection efficiency increases with the increase of thickness, and reaches the maximum value at the critical thickness and thereafter found to gradually decrease. Due to the self-absorption effect of the conversion material, the intrinsic detection efficiency of the thin-film-coated SiC neutron detector is limited to less than 5%. Fortunately, the trench type SiC neutron detector can break through this limitation. However, under current process conditions, SiC can only be used for shallow trench etching. In order to increase the neutron detection efficiency as high as possible in the limited trench depth, we have proposed a scheme of filling the trench with a double conversion layer, which can significantly improve the intrinsic detection efficiency of neutrons compared with the traditional trench structure with a single conversion layer. Apart from estimating the intrinsic detection efficiency, the energy deposition spectrum in the SiC detector region by the produced secondary charged particles upon thermal neutron interaction with conversion material(10B and 6LiF) have also been researched for various structural parameters such as conversion layer thickness, trench width and trench gap of detector.

The aim of the NEPTUNE (Nuclear process-driven Enhancement of Proton Therapy UNravEled) project is to investigate in detail both the physical and radiobiological phenomena that could justify an increase of the proton-induced cytogenetic effects in cells irradiated in presence of an agent containing natural boron. In this work, a double-stage silicon telescope coupled to different boron converters was irradiated at the CATANA proton therapy facility (INFN-LNS) for studying the proton boron fusion and the neutron boron capture reactions by discriminating secondary particles from primary protons. Different boron targets were developed by depositing boric acid, enriched with a higher than 99% content of 10B or 11B, on a 50 µm thick PolyMethilMetacrylate (PMMA) substrate. The 10B target allows to evaluate the contribution of lithium and alpha particles produced by the boron neutron capture reaction triggered by secondary thermal neutrons, while the 11B target is exploited for studying the effect of the p + 11B → 3α nuclear reaction directly triggered by primary protons. Experimental results clearly show the presence of alpha particles from both the reactions. The silicon telescope is capable of discriminating, by means of the so-called “scatter plots”, the contribution of alpha particles originated by thermal neutrons on 10B with respect to the ones produced by protons impinging on 11B. Although a reliable quantitative study of the alpha production rate has not been achieved yet, this work demonstrates that low energy and, therefore, high-LET particles from both the reactions can be measured.

In this study, 12 trace elements were measured in the shell and muscles of two species of marine and farmed Jinga shrimp Metapenaeus ensis and Litopenaeus vannamei in the Persian Gulf zone. For this purpose, the instrumental neutron activation analysis and gamma- ray spectrometry were used. The Isfahan Miniature Neutron Source Reactor as the source of thermal neutrons and high purity germanium detector for gamma-ray spectrometry was used. The aim of this study was to determine useful and unnecessary elements in farmed and marine shrimps in Persian Gulf region. The results show that the concentrations of aluminum, calcium, magnesium, manganese, bromine, and scandium in the shell samples are significantly higher than in the muscles samples. Also, shrimp muscles are a good source of useful elements such as zinc, magnesium, and manganese. And the amount of arsenic and aluminum in it is negligible.

We test whether X-ray micro-computed tomography (microCT) imaging can be used as a tool for screening magnetite grains to improve the accuracy and precision of cosmogenic 3 He exposure dating. We extracted detrital magnetite from a soil developed on a fanglomerate at Whitewater, California, which was offset by the Banning strand of the San Andreas Fault. This study shows that microCT screening can distinguish between inclusion-free magnetite and magnetite with fluid or common solid inclusions. Such inclusions can produce bulk 3 He concentrations that are significantly in excess of the expected spallation production. We present Li concentrations, major and trace element analyses, and estimated magnetite (U–Th) / He cooling ages of samples in order to model the contribution from fissiogenic, nucleogenic, and cosmogenic thermal neutron production of 3 He. We show that mineral inclusions in magnetite can produce 3 He concentrations of up to 4 times that of the spallation component, leading to erroneous exposure ages. Therefore, grains with inclusions must be avoided in order to facilitate accurate and precise magnetite 3 He exposure dating. Around 30 % of all grains were found to be without inclusions, as detectable by microCT, with the largest proportion of suitable grains in the grain size range of 400–800 µ m. While grains with inclusions have 3 He concentrations far in excess of the values expected from existing 10 Be and 26 Al data in quartz at the Whitewater site, magnetite grains without inclusions have concentrations close to the predicted depth profile. We measured 3 He concentrations in aliquots without inclusions and corrected them for Li-produced components. By comparing these data to the known exposure age of 53.5 ± 2.2 ka, we calibrate a production rate for magnetite 3 He at sea level and high latitude (SLHL) of 116 ± 13 at g −1 a −1 . We suggest that this microCT screening approach can be used to improve the quality of cosmogenic 3 He measurements of magnetite and other opaque mineral phases for exposure age and detrital studies.

—A technique for studying the resistance of IR-wavelength LEDs to the action of fast neutrons at the IRT-T reactor is described. A thermal neutron filter based on boron carbide and cadmium has been developed for the GEC-6 horizontal experimental channel of the IRT-T reactor. The identity of degradation processes in LEDs is shown when exposed to fast neutrons with pulsed (BARS-4 reactor) and continuous (GEC-6 of IRT-T reactor) fluence. The developed technique can be recommended for studying the resistance of various materials and electronic products to the action of fast neutrons in various irradiation regimes.

A laboratory technique was developed for obtaining an ultra-low-background hybrid material based on organic glass — polymethyl methacrylate (PMMA). 157Gd nuclei are used as an efficient absorber of thermal neutrons in the hybrid material. A uniform distribution of gadolinium in the PMMAmatrix is achieved by introducing the gadolinium in the form of a coordination compound — gadolinium acetylacetonate — into the hybrid material. A uniform gadolinium distribution was achieved in a 5 cm thick matrix of hybrid material with gadolinium mass content from 1.0 to 1.5%. It is shown that at 298 K the mechanical characteristics of the hybrid material samples coincide with nominally pure PMMA, but at 77 K the hybrid material is inferior to pure PMMA within admissible requirements.

Applications of neutron-gamma analysis for chlorine determination in different materials (e.g., solids, aqueous solutions, and surface contaminations) are discussed. Solid samples were measured with a pulsed fast/thermal neutron analysis (PFTNA) system consisting of a pulsed neutron generator and sodium iodide gamma detectors and an associated particle imaging (API) system consisting of a neutron generator (with a built-in alpha detector) and sodium iodide gamma detector; all sample measurements were made under similar conditions (neutron flux values and detector-side area for crossing gamma flux). The advantages of PFTNA over the API method were demonstrated based on comparison of metrological parameters (i.e., minimal detectible level for a 1-h measurement period and clock time required to reach 5% measurement accuracy). For these reasons, the PFTNA system in scanning mode effectively identified areas of artificial chlorine contamination in the field. Paired with GPS, a chlorine contamination map was created from PFTNA scanning results; this map accurately identified the actual position and size of chlorine contaminated locations in the field. Chlorine-contaminated areas of ~4.5 m2 were effectively located within the ~500 m2 field during a ~2-h scanning period. Based on the results from this work, promising directions for further application of the PFTNA method are disclosure of chlorine contamination in soil and measurements of marine water salinity.

Organoboron compounds have been playing an increasingly important role in analytical chemistry, material science, health applications, and particularly as functional polymers like boron carriers for cancer therapy. There are two main applications of boron isotopes in radiation cancer therapy, Boron Neutron Capture Therapy and Proton Boron Fusion Therapy. In this study, a novel and original material consisting of a three-dimensional polymer network crosslinked with $$^{10}$$ 10 B enriched boric acid molecules is proposed and synthesized. The effects of the exposition to thermal neutrons were studied analyzing changes in the mechanical properties of the proposed material. Dedicated Monte Carlo simulations, based on MCNP and FLUKA main codes, were performed to characterize interactions of the proposed material with neutrons, photons, and charged particles typically present in mixed fields in nuclear reactor irradiations. Experimental results and Monte Carlo simulations were in agreement, thus justifying further studies of this promising material.

A description is given of results from developing a new lithium-loaded plastic scintillator based on a styrene–methacrylic acid copolymer. Measurements of the light output and transparency of samples of the new material allow the optimum concentrations of the scintillation additive, wave shifter, secondary solvent, and lithium-containing additive to be determined.

Recent advances in semiconductor processing technology have enabled development of a solid-state neutron sensor with exceptional efficiency for its unique reduced size. The boron-10 ( $n$ , alpha) capture reaction is used to detect incident thermal neutrons. This reaction offers significant advantages for a small detector because the reaction products have high energies which can generate a large number of charge carriers in semiconductor materials by impact ionization. The enabling and differentiating technology of the sensor design is a stack consisting of multiple bilayers alternating between converter material (boron) and collector semiconductor material (silicon). As the number of these bilayers increases, the probability of a neutron interacting with boron-10 in at least one of the bilayers—so it is captured in the detector before it has a chance to pass through—approaches unity. The sensor can be made very thin while also remaining highly efficient, offering a unique possibility for low-profile in situ dosimetry. Calculations have shown that a multilayer sensor with an efficiency greater than 70% is feasible in a sensor stack 2 mm thick.

The impact of chromium oxide (Cr2O3) on the organizational, physical, mechanical, and photon shielding qualities of new chromium borate glasses with the chemical composition 50B2O3-25Bi2O3-25Li2O3-xCr2O3 (x=0 (Cr-00), 0.2 (Cr-0.2), 0.4 (Cr-0.4), 0.8 (Cr-0.8), and 1.5 (Cr-1.5) wt%) is investigated. The amorphous behavior of these glasses was verified via X-ray diffraction (XRD) and scan electron microscope (SEM) micrographs. Fourier transform infrared (FTIR) spectra of prepared glass samples revealed the presence of absorption bands due to Cr3+ ions. The density (ρglass) increased as the Cr2O3 concentration increased. The mass attenuation coefficients (MACs) of the proposed glasses were evaluated via the FLUKA code, XCOM program, and Phy-X/PSD software. The MACs values followed this trend: (MAC)Cr-00 (MAC,MFP)Cr-0.2 > (MAC,MFP)Cr-0.4 > (MAC,MFP)Cr-0.8 > (MAC,MFP)Cr-1.5. The Cr-00 glass sample had the lowest effective atomic number (Zeff) values, while the Cr-1.5 glass sample had the highest. The calculated exposure buildup factor (EBF) values decreased as the Cr2O3 content increased. The Cr-1.5 sample had a higher effective removal cross-section for fast neutrons (ΣR) value than other radiation shielding materials. The addition of Cr2O3 increased the rigidity and elastic constants (Young’s (Y), bulk (K), shear (G), and longitudinal (L) moduli) and improved the level of radiation shielding provided. The prepared glasses can be used as gamma and neutron shielding tools in the medical area.

The IRT-T reactor has been conducting research in the field of irradiation of ingots of single-crystal semiconductor materials since 1987. The article describes the existing silicon doping facility. The results of studies on the possibility of creating an additional irradiation channel for neutron-transmutation doping of silicon are presented. It is shown that the use of a graphite reflector and a thermal neutron filter based on boron makes it possible to achieve non-uniformity of irradiation up to 5 %. The principal possibility of irradiating single-crystal silicon ingots with a diameter of up to 203 mm and a length of up to 500 mm is shown. The questions of optimizing the configuration of the core and the regime of reactors operation for increasing the neutron flux in the irradiation channels are discussed. In addition, applying the facility to produce base materials for neutron dosimeter in neutron capture therapy studies is proposed.

In this study, the neutron shielding and mechanical properties of short carbon fiber reinforced aluminium 6061-boron carbide hybrid composite have been investigated by simulation and experimental methods. The simulation results indicated that in the energy range of the epithermal neutrons which have greater kinetic energy than thermal neutrons (0.0253ev), the introduction of fiber materials containing carbon atoms into aluminium 6061-boron carbide increases the probability of neutron scattering, meanwhile the epithermal neutrons slowed down by the carbon atoms, hence the shielding performance will be improved. Additionally, compared with aluminium 6061-boron carbide composite, it was found that the short carbon fiber reinforced hybrid composite has better strengthening effect on matrix, mechanical tests show the enhancement in tensile strength and hardness. Moreover, the short carbon fiber can hinder crack propagation and absorb fracture work during the stress process while enhance the toughness of the neutron shielding composite.

Measurements and simulations of thermal neutron attenuation by printed circuit boards are compared. Attenuation coefficients in typical epoxy-resin/glass-fibre substrate material can be as high as 2 cm-1, corresponding to 27% attenuation by 1.6 mm of substrate. Attenuation is attributed to neutron scattering off hydrogen in the resin acting in synergy with absorption by boron in the glass; this effect is substantially greater than that estimated from absorption by boron alone. Design of thermal neutron detector assemblies should take this attenuation into account and may require board thickness to be minimised or specialised substrate materials to be used.

Due to high thermal neutron absorption cross-section, boron (B) content is a binding specification for nuclear grade graphite. Reliable analytical methods are essential to determine boron at trace to ultra-trace levels. Analytical Chemistry Division, Bhabha Atomic Research Centre, India has prepared an in-house graphite reference material (GRM). A Charged particle activation analysis methodology has been developed for quantification of B in graphite matrix. Detection limit for boron using this methodology is 1.4 mg kg−1. This in-house GRM can be used for quantification of B in unknown graphite samples, validating existing analytical methods and for the development of new analytical methods.

The last decade witnessed a vast advancement in several fields of science and technology, including 3D printing. 3D printing is utilized in a wide range of applications, from printing small objects to full-size houses. This research aims to examine thermal neutron attenuation using Acrylonitrile Butadiene Styrene (ABS) filament infused with a thermal neutron absorbent, employing a Monte Carlo simulation toolkit. Several thermal neutron absorbers were tested individually (gadolinium, boron, gold, and cadmium). The simulation results showed a two-centimeter-thick slab infused with boron reduced the total relative dose to one half. Thus, this sample has the potential to be an excellent thermal neutron shielding material. A four-centimeter-thick slab infused with boron registered a neutron relative dose reduction like that of water, which makes the sample a good candidate for water equivalent phantom manufacturing. As for the gadolinium-infused slab, the results indicate that it could be used as a detector cover to discriminate against neutrons in a mixed radiation field. Infusing the ABS slab with gold reduced gamma-ray dose to around 40% of the total relative dose compared to the bare ABS filament. The neutron relative dose remained unchanged for the gold-infused ABS filament. The use of ABS slab infused with cadmium resulted in a steady reduction of the neutron relative dose compared to the bare ABS filament, while the gamma-ray dose remained unchanged. The results for the ABS slab infused with gold and cadmium indicate that these filaments can be used for radiation detection purposes in a mixed neutron-gamma field.

We report γ − ray shielding effectiveness of eight selected shielding materials susceptible to be used for the conception of γ − ray and fast neutrons absorbers has been investigated. Four of the selected materials were polyethylene composites, the first one is doped with 7.5% Lithium (Lithium7.5%-Polyethylene), the second one is doped with 8.97% boron (Borated8.97%-Polyethylene), the third one is doped with 5.45% boron (Borated5.45%-Polyethylene), and the fourth one contains high hydrogen content as well as 0.9% by weight boron namely Self Extinguishing Borated Polythylene. The other four considered materials are FlexiBoron, High Temperature Boron Silicone, Polykast Dry Mix, and Field castable Heat Resistant shielding. These materials have been selected in fact because they contain iron (Fe), zinc (Zn), Strontium (Sr) and Manganese (Mn) which heavy elements and considered as good absorbers of γ − radiation and also because they contain hydrogen which is considered as the furthermost active moderator as well as boron which is an extremely appropriate absorber of thermal neutrons. The study has been conducted by determining the γ − ray energy absorption buildup factors (EABF) as well as the exposure buildup factors (EBF). The evaluation of the buildup factors has been determined by employing the five parameters geometric progression (G-P) fitting approximation and taking into account the reference standard database (1991-ANSI/ANS-6.4.3). This determination has been carried out using a photon energy varying from 0.015 to 15 MeV with a permeation depth until 40 mean free paths. By investigating on the reliance of EABF and EBF on the incident photon energy, chemical arrangement, and permeation depth, a deeply comparison between the obtained values of EABF and EBF has been performed. By inspecting the obtained values of EABF and EBF established for the first time, our numerical simulations allow to conclude that among the selected shielding materials, Field castable Heat Resistant shielding and 5.45% Borated Polyethylene provide the lower buildup factors at lower energies ( 1.5 MeV for EABF, 2 MeV for EBF), suggesting thereby their candidacy to serve as potential shielding materials against γ − radiation in this energy range. The results of the present investigation therefore shed light on selectivity of the appropriate shielding material as well as the determination of the optimum thickness of the shielding material subject of having superior neutrons shielding ability.

A Monte Carlo model of the thermal-neutron scintillation detector based on NE 912 lithium glass has been created and verified. The simulation was validated by comparing its result to experimental data from a prototype detector exposed to thermal-neutron and γ-ray beams. The light yield in the scintillator for a captured thermal neutron, the quenching factor, and the decay times of the scintillator were determined. The accuracy in reproducing the pulse shapes obtained in the experiment is sufficient to allow analysis of experimental data and estimation of the efficiency of n/γ discrimination techniques. Based on the simulation, it is possible to develop detector models with a low γ-ray sensitivity using heterogeneous composite scintillators with various geometries.

No thermodynamic data for the solution of Cl in ferrous alloys were found in the literature. This is in accord with recent Accelerator Mass Spectroscopy (AMS) analyses which showed that Cl contents in stainless steel (SS) are in the order of a few ppb. However, based on older chemical analyses of Cl in the order of 100 ppm, SS that has been irradiated with thermal neutrons in nuclear reactors is considered a major source of the long-lived 36Cl isotope in nuclear waste. In this study, the potential Cl contamination of SS originating from production and refinement processes is investigated. Unlike ferrous alloys, blast-furnace and steelmaking slags can dissolve significant amounts of Cl. The equilibrium distribution of Cl species between slags and gas phase was calculated for various steelmaking processes using the FactSage 7.2 software and databases. The results showed that despite the high volatility of metal chlorides at high temperatures, significant fractions of Cl can be retained in the slag phase even at 1600 °C. Chloride may also be incorporated in non-metallic inclusions originating from secondary refining. Based on these results and on several further assumptions, various scenarios for explaining, and also avoiding, Cl contamination of steel are discussed.

The IBR-2 (JINR) pulsed reactor is used in the field of solid state physics for experiments on extracted beams. It is predicted that the IBR-2 reactor will reach resource limits in 2032–2037. To keep and develop beam research at JINR, the development of a new pulsed neutron source has begun. According to estimates, the reactor will have average thermal power 12–15 MW, extracted by sodium coolant. At this power, the average flux density of thermal neutrons on the surface of the water moderator will reach 1014 sec–1·cm–2 and the peak density 5·1016 sec–1·cm–2, which is more than 10 times higher than the analogous parameters of the operating IBR-2 reactor. The new reactor is supposed to use neptunium-based fuel. The main characteristics and concept of the new reactor are presented.

The Kansai BNCT Medical Center has a cyclotron based epithermal neutron source for clinical Boron Neutron Capture Therapy. The system accelerates a proton to an energy of 30 MeV which strikes a beryllium target producing fast neutrons which are moderated down to epithermal neutrons for BNCT use. While clinical studies in the past have shown BNCT to be highly effective for malignant melanoma of the skin, to apply BNCT for superficial lesions using this system it is necessary to shift the thermal neutron distribution so that the maximum dose occurs near the surface. A dose distribution shifter was designed to fit inside the collimator to further moderate the neutrons to increase the surface dose and reduce the dose to the underlying normal tissue. Pure polyethylene was selected, and a Monte Carlo simulation was performed to determine the optimum thickness of the polyethylene slab. Compared with the original neutron beam, the shifter increased the thermal neutron flux at the skin by approximately 4 times. The measured and simulated central axis depth distribution and off axis distribution of the thermal neutron flux were found to be in good agreement. Compared with a 2 cm thick water equivalent bolus, a 26% increase in the thermal neutron flux at the surface was obtained, which would reduce the treatment time by approximately 29%. The DDS is a safe, simple and an effective tool for the treatment of superficial tumours for BNCT if an initially fast neutron beam requires moderation to maximise the thermal neutron flux at the tissue surface.

An electrodeposited thorium target and a ThO2 sample were irradiated together with fast neutrons using the Chinese Fast Burst Reactor II. The activities of 233Pa in both samples were measured and c...

Yttrium hydride is being considered as a moderator material for microreactor concepts because of its excellent hydrogen retainment capacity at high temperatures. These types of reactors, operating ...

The neutron capture cross sections of 244Cm and 246Cm were measured for the neutron energy range of 1–1000 eV via the neutron time-of-flight method with Accurate Neutron-Nucleus Reaction measurem...

Background/Aim: Tumor cell destruction by boron neutron capture therapy (BNCT) is attributed to the nuclear reaction between (10)B and thermal neutrons. The accumulation of (10)B atoms in tumor cells without affecting adjacent healthy cells is crucial for effective BNCT. We previously reported that several types of liposomal boron delivery systems (BDS) delivered effective numbers of boron atoms to cancer tissues, and showed tumor-growth suppression after thermal neutron irradiation. In the present study, we examined the effects of BNCT after intra-arterial infusion of (10)B-borono-dodecaborate ((10)BSH) by liposomal BDS in rabbit hepatic cancer models. Materials and Methods: We prepared (10)BSH-entrapped transferrin-conjugated polyethylene glycol liposomes constructed with distearoyl-boron lipid (TF-PEG-DSBL), and performed thermal neutron irradiation at the Kyoto University Institute for Integrated Radiation and Nuclear Science after intra-arterial infusion into rabbit VX-2 hepatic tumors. Results: Concentrations of (10)B in VX-2 tumors on delivery with TF-PEG-DSBL liposomes reached 25 ppm on day 3 after the injection. Tumor growth was suppressed by thermal neutron irradiation after intra-arterial injection of this (10)BSH-containing liposomal BDS, without damage to normal cells. Conclusion: The present results demonstrate the applicability of (10)B-containing TF-PEG-DSBL liposomes as a novel intra-arterial boron carrier in BNCT for cancer.

In this study, the fast neutron and gamma-ray absorption capacities of the new glasses have been investigated, which are obtained by doping CoO,CdWO4,Bi2O3, Cr2O3, ZnO, LiF,B2O3 and PbO compounds to SiO2 based glasses. GEANT4 and FLUKA Monte Carlo simulation codes have been used in the planning of the samples. The glasses were produced using a well-known melt-quenching technique. The effective neutron removal cross-sections, mean free paths, half-value layer, and transmission numbers of the fabricated glasses have been calculated through both GEANT4 and FLUKA Monte Carlo simulation codes. Experimental neutron absorbed dose measurements have been carried out. It was found that GS4 glass has the best neutron protection capacity among the produced glasses. In addition to neutron shielding properties, the gamma-ray attenuation capacities, were calculated using newly developed Phy-X/PSD software. The gamma-ray shielding properties of GS1 and GS2 are found to be equivalent to Pb-based glass.

Scattering of thermal neutrons and Doppler broadening of epithermal neutron resonances in uranium and its compounds may be sensitive to crystal binding. The thermal scattering law (TSL) for uranium...

The Bonner sphere technique with activation foil made of natural tantalum is used to estimate doses during secondary particle irradiation. The neutron spectrum in the isocenter of a Varian Trilogy medical linear accelerator (linac) is calculated. The average neutron energy and the effective neutron cross section are determined. The flux of the (γ, n) reaction is calculated using a measured spectrum. The fluxes measured using Bonner spheres with a natural tantalum activation target correspond to the results from similar measurements obtained in other ways.

The cross sections of 89Y(n, 3n)87m,87gY and 209Bi(n, 4n)206Bi reactions at a neutron energy of 30 MeV are measured by making use of neutron beams of continuous energy spectra and a subtraction method. By impinging proton beams of 30 and 35 MeV to a thick beryllium target, neutron beams of continuous and broad energy spectra are produced and are guided to Y and Bi sample targets. The difference between the two neutron spectra generated by two neighboring proton energies is found to be peaked in a narrow energy range and thus can be regarded as quasi-monoenergetic, which can be used to extract (n, xn) cross sections. The uncertainty in the neutron fluence is reduced by analyzing the activities of aluminum and niobium reference samples placed on top of the Y and Bi samples. The use of a subtraction method by employing neutron beams of continuous energy spectra gives us the 89Y(n, 3n)87m,87gY and 209Bi(n, 4n)206Bi cross sections in fair agreement with the existing experimental data and nuclear data libraries.

The present work was aimed to show the possibility of using hydrogel (acrylic/boric acid) for evaluation of the neutron radiation shielding. The influence of acrylic acid concentration, different gamma doses and relative contents of boric acid were studied. The physical properties and the thermomechanical stability of the studied samples were investigated. The shielding property of the composite for neutron was tested by Pu–Be neutron source (5 Ci) under room temperature. The neutron fluence rates and gamma fluxes were measured using a stilbene organic scintillator. The macroscopic effective removal cross-section ΣR (cm−1) of fast neutrons and total attenuation coefficient μ (cm−1) of gamma rays has been studied experimentally. The transmission parameters, the relaxation length (??) and the half-value layer (HVL) were obtained. The obtained results indicated that the addition of boric acid to acrylic acid tends to increase the macroscopic effective removal cross-section ΣR (cm−1) to 0.141 compared to 0.094 of ordinary concrete.

Instrumental neutron activation analysis may need to correct for matrix interferences caused by fast neutron induced threshold reactions. The very low Al mass fraction in bovine liver Standard Reference Material (SRM) has eluded certification due to such complications. Cold neutron beam irradiation absent fast neutrons provides a possible alternative, which has been applied for the first time to determine Al in bovine liver SRM 1577c. The Al mass fraction was determined to be (0.80 ± 0.15) mg/kg using the traditional single peak-fitting method, and (0.78 ± 0.23) mg/kg using a new physics model-based spectral fitting method.

The neutron temperature at different irradiation positions in the research reactor FRM II was measured using thermometer labels together with lutetium standards. This was the first time to measure the local temperature at the irradiation positions with this combination. The simplified Westcott formalism was used to calculate the neutron temperature using Lu and g(Tn) factors from different data sources. The results showed agreement between neutron temperature according to the Lu method and the temperature showed on the labels. They can be used as a reliable alternative to determine the neutron temperature instead of using Lu standards in the extended k0 NAA.

Impacts of lead and bismuth oxides insertion on a novel glass system of P2O5, B2O3, Li2O, Al2O3 according the formula 25B2O3–25P2O5–10Li2O–5Al2O3–5ZnO-xPbO+ (30-x)Bi2O3, x = 5,10, 15, 20, 25 mol%. The mass attenuation coefficient (μm = μ/ρ) simulated between 0.015 and 15 MeV using MCNP and calculated theoretically using Phy-X/PSD program. Based on the simulated μm, other significant parameters such as linear attenuation coefficient (LAC), half and tenth value layer (HVL, TVL), mean free path (MFP), and effective atomic number (Zeff) were calculated for fabricated glasses. The G-P fitting methods were used to calculate the exposure and energy absorption buildup factors (EBF and EABF) for fabricated glasses. Furthermore, fast neutron removal cross sections (ΣR) were calculated theoretically for fabricated glasses. The prepared glasses were effective shielding material which can reduce fast neutrons as well as gamma rays.

The production of a deuteron beam in a vacuum-insulated tandem accelerator and the generation of fast neutrons from a lithium target have been investigated. The possibility of using a source of fast neutrons for radiation testing of materials and for fast-neutron therapy is discussed.

Введение. Бор-нейтронозахватная терапия (БНЗТ) является перспективной экспериментальной методикой лечения онкологических заболеваний. По данным клинических исследований пациентов с глиобластомой и меланомой, леченных БНЗТ на ядерных реакторах, отмечены рост медианы выживаемости и улучшение качества жизни. Для получения эпитепловых нейтронов ведется разработка новых источников на основе ускорителей заряженных частиц. Один из проектов был реализован в Институте ядерной физики им. Г.И. Будкера СО РАН, достигнутые параметры пучка позволяют проводить доклинические исследования. Цель работы - выявление на клеточных линиях глиомы и меланомы зависимости эффективности БНЗТ от концентрации бора при использовании пучка, генерируемого на источнике эпитепловых нейтронов ускорительного типа Института ядерной физики им. Г.И. Будкера СО РАН и оценка перспектив использования этого источника эпитепловых нейтронов для дальнейших клинических исследований. Методика. Клеточные линии U251 (глиома) и SK-Mel28 (меланома) опухолей человека инкубировали с добавлением в ростовую среду различных концентраций бора, используя препарат борфенилаланин, в течение 24 ч, затем облучали потоком нейтронов. Измерение концентрации изотопа бора 10В в опухолевых клетках проводили на атомно-эмиссионном спектрометре ICPE-9820 (Shimadzu, Япония). Клоногенный тест использовали для оценки влияния бор-нейтронозахватной терапии на клетки глиомы и меланомы. Результаты. Анализ данных БНЗТ показал, что колониеобразующие свойства облученных клеток глиомы и меланомы уменьшались с повышением концентрации бора. Так, по мере накопления бора линией SK-Mel28 увеличивается количество погибших клеток после облучения, а концентрация 10В 25 мкг/мл обеспечивает летальную дозу для 100% клеток (LD100). Глиальная линия накапливает бор менее интенсивно и гибель 100% клеток происходит при концентрации 10В 50 мкг/мл. В образцах, облученных без бора, в сравнении с контролем также наблюдалось снижение выживаемости клеток из-за присутствия быстрых нейтронов и гамма-излучения. Заключение. Данные экспериментов in vitro доказывают эффективность действия бор-нейтронозахватной терапии на клетки глиомы и меланомы при использовании источника эпитепловых нейтронов ускорительного типа ИЯФ СО РАН и борфенилаланина как агента доставки бора с концентрацией 10В 6,25-50 мкг/мл, а также перспективность использования данного метода в лечении таких опухолей, как глиома и меланома. Boron neutron capture therapy (BNCT) is a promising experimental method for the treatment of oncological diseases. According to results of clinical trials, patients with glioblastoma and melanoma treated with BNCT at nuclear reactors showed an increase in median overall survival and an improvement in quality of life. To obtain epithermal neutrons, new sources based on charged particle accelerators are being developed. One of the projects was implemented at the G.I. Budker Institute of Nuclear Physics, and the obtained beam parameters allowed conducting preclinical experiments. The aims were to identify the dependence of the effectiveness of BNCT in glioma and melanoma cell lines on boron concentrations using a beam generated at the accelerator based epithermal neutron source in the G.I. Budker Institute of Nuclear Physics and to evaluate prospects for using this epithermal neutron source for further clinical research. Methods. The U251 glioma cell line and the SK-Mel28 melanoma cell line were incubated with various concentrations of boronophenylalanine added to the growth medium for 24 hours and then irradiated with a neutron flux. The 10B accumulation in tumor cells was measured with an ICPE-9820 atomic emission spectrometer (Shimadzu, Japan). The effect of BNCT on glioma and melanoma cells was evaluated by the colony forming assay. Results. Analysis of the BNCT experimental data showed that the colony-forming capabilities of irradiated glioma and melanoma cells decreased in proportion to the increase in boron concentration. Thus, increasing accumulation of boron by SK-Mel 28 cells provided a greater number of dead cells with irradiation at a concentration of 10B of 25 µg/ml being a lethal dose for 100% of the cells (LD100). The glial cell line accumulated boron less intensively; death of 100% of cells occurred at a 10B concentration of 50 µg/ml. In samples irradiated without boron, the number of colonies was also decreased compared to the control due to the presence of fast neutrons and gamma-radiation components. All differences between the control and the experiment were statistically significant (p

This paper details and implements a framework for evaluating thermal neutron scattering cross sections that provide S ( α , β ) data and covariance data for hydrogen in light water. This methodolog...

Magnesium oxide (MgO) and Zinc oxide (ZnO) nanoparticles (NPs) have been successfully synthesized by solid–solid reaction method. The structural properties of ZnO and MgO NPs were studied using X-ray diffraction (XRD) and scanning electron microscopy (SEM). XRD results indicated a formation of pure MgO and ZnO NPs. The mean diameter values of the agglomerated particles were around to be 70 and 50 nm for MgO and ZnO NPs, respectively using SEM analysis. Further, a wide-range of nuclear radiation shielding investigation for gamma-ray and fast neutrons have been studied for Magnesium oxide (MgO) and Zinc oxide (ZnO) samples. FLUKA and Microshield codes have been employed for the determination of mass attenuation coefficients (μm) and transmission factors (TF) of Magnesium oxide (MgO) and Zinc oxide (ZnO) samples. The calculated values for mass attenuation coefficients (μm) were utilized to determine other vital shielding properties against gamma-ray radiation. Moreover, the results showed that Zinc oxide (ZnO) nanoparticles with the lowest diameter value as 50 nm had a satisfactory capacity in nuclear radiation shielding. © 2020 Korean Nuclear Society

The article presents an experimental confirmation of the operability of neutron concentrators in devices that form and use directed high-intensity thermal neutron beams with elliptical channels made as blocks of profiled graphite and aluminum plates. The effect of neutron reflection from the surface of materials is the basis of a device capable of selecting neutrons by their directions in space. The study experimentally confirmed the efficiency of a moderating-focusing structure (MFS) based on a pack of elliptical neutron mirrors, which makes it possible to form oriented thermal neutron beams from the outgoing neutron flux. To record the effects of selective thermal neutron separation, silicon single-crystal wafers were used, due to which it was possible to obtain portraits of integral neutron fluxes in the reactor. The experiments were carried out in a horizontal experimental channel (HEC-4) at the IRT-T reactor of the National Research Tomsk Polytechnic University. The integral neutron flux was (2.3–3.02)·1017 cm–2. The neutron flux was detected by the change in the specific electrical resistivity of the single-crystal silicon wafers. The effect of concentration of thermal neutrons was recorded both on the block of graphite neutron mirrors and on the block of aluminum thin-walled elliptical mirrors. In the near future, on this basis, it will be possible to solve such problems as extending the reactor life by reducing the hydrogen uptake in the inner walls. In addition, the experiments have proved the possibility of creating anisotropic structures that lie outside the formalism of Liouville’s theorem, in which the surfaces of thermal neutron sinks are formed with subsequent concentration in the areas separated by aluminum or graphite plates.