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In this research, curium oxide (CmO2) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO2 is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO2 and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO2 is competent like UO2 as a fuel due to higher fission cross-section of 243Cm and 245Cm isotopes and neutron capture cross-section of 244Cm and 246Cm isotopes. As a result, when some or all of the UO2 of fuel rods or IBA rods are replaced by CmO2, we get a similar k-inf like the reference even with lower enrichment UO2 fuels. These studies show that the use of CmO2 as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO2 with inert material ZrO2 in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

Neutron capture cross section measurements in time-of-flight facilities are usually performed by detecting the prompt γ-rays emitted in the capture reactions. One of the difficulties to be addressed in these measurements is that the emitted γ-rays may change with the neutron energy, and therefore also the detection efficiency. To deal with this situation, many measurements use the so called Total Energy Detection (TED) technique, usually in combination with the Pulse Height Weighting Technique (PHWT). With it, it is sought that the detection efficiency depends only on the total energy of the γ-ray cascade, which does not vary much with the neutron energy. This technique was developed in the 1960s and has been used in many neutron capture experiments to date. One of the requirements of the technique is that γ-ray detectors have a low efficiency. This has meant that the PHWT has been used with experimental setups with low detection efficiencies. However, this condition does not have to be fulfilled by the experimental system as a whole. The main goal of this work is to show that it is possible to measure with a high efficiency detection system that uses the PHWT, and how to analyze the measured data., This work was supported in part by the I+D+i grant PGC2018- 096717-B-C21 funded by MCIN/AEI/10.13039/501100011033 and by the European Commission H2020 Framework Programme project SANDA (Grant agreement ID: 847552).

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.]

This work addresses the neutronic design of a Beam Shaping Assembly (BSA) that is appropriate for a Accelerator-based Boron Neutron Capture Therapy (AB-BNCT). Instead of designing a moderator utilizing the IAEA-1223 recommendation and in-phantom figure of merit, in this work we study neutronics to decrease the out-of-beam dosimetry by employing radiobiological figures of merit to assess the beam quality. In particular, in the radiobiological probability model, we include two normal tissue cases, and accordingly optimize the thickness and cone angle of the collimator. Hence, in this study, for a large opening diameter, i.e., 10 cm, we use a small cone angle, which can decrease the out-of-beam dosimetry effectively. Accordingly, for a small opening diameter, i.e., 2 cm in this study, we employ a large cone angle to guarantee that the neutron flux is large enough. Our simulation results demonstrate that the neutronics optimization of the collimator is a tradeoff between the out-of-beam dosimetry and the neutron flux. From our findings, we conclude that upon considering radiobiological figures of merit, collimators with a tunable design shape are more appealing due to adapting to different treatment situations.

In this study, molecular interactions of prostate-specific membrane antigen (PSMA) with five chemically distinct urea-based boron-containing inhibitors have been investigated at the atomic level using molecular docking and molecular dynamics simulations. The PSMA–inhibitor complexations have been analyzed by comparing their binding modes, secondary structures, root-mean-square deviations, noncovalent interactions, principal components, and binding free energies. PSMA is a cell surface glycoprotein upregulated in cancerous cells and can be targeted by boron-labeled inhibitors for boron neutron capture therapy (BNCT). The effective BNCT requires the selective boron delivery to the tumor area and highly specific PSMA-mediated cellular uptake by tumor. Thus, a potent inhibitor must exhibit both high binding affinity and high boron density. The computational results suggest that the chemical nature of inhibitors affects the binding mode and their association with PSMA is primarily dominated by hydrogen bonding, salt bridge, electrostatic, and π–π interactions. The binding free energies (−28.0, −15.2, −43.9, −23.2, and −38.2 kcal/mol) calculated using λ-dynamics for all inhibitors (In1–5) predict preferential binding that is in accordance with experimental data. Among all inhibitors, In5 was found to be the best candidate for BNCT. The binding of this inhibitor to PSMA preserved its overall secondary structure. These results provide computational insights into the coordination flexibility of PSMA and its interaction with various inhibitors. They can be used for the design and synthesis of efficient BNCT agents with improved drug selectivity and high boron percentage.

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.

Introduction Boron neutron capture therapy (BNCT) is a biologically targeted, cell-selective particle irradiation therapy that utilizes the nuclear capture reaction of boron and neutron. Recently, accelerator neutron generators have been used in clinical settings, and expectations for developing new boron compounds are growing. Methods and Results In this study, we focused on serum albumin, a well-known drug delivery system, and developed maleimide-functionalized closo-dodecaborate albumin conjugate (MID-AC) as a boron carrying system for BNCT. Our biodistribution experiment involved F98 glioma-bearing rat brain tumor models systemically administered with MID-AC and demonstrated accumulation and long retention of boron. Our BNCT study with MID-AC observed statistically significant prolongation of the survival rate compared to the control groups, with results comparable to BNCT study with boronophenylalanine (BPA) which is the standard use of in clinical settings. Each median survival time was as follows: untreated control group; 24.5 days, neutron-irradiated control group; 24.5 days, neutron irradiation following 2.5 h after termination of intravenous administration (i.v.) of BPA; 31.5 days, and neutron irradiation following 2.5 or 24 h after termination of i.v. of MID-AC; 33.5 or 33.0 days, respectively. The biological effectiveness factor of MID-AC for F98 rat glioma was estimated based on these survival times and found to be higher to 12. This tendency was confirmed in BNCT 24 h after MID-AC administration. Conclusion MID-AC induces an efficient boron neutron capture reaction because the albumin contained in MID-AC is retained in the tumor and has a considerable potential to become an effective delivery system for BNCT in treating high-grade gliomas.

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.

Neutron capture yields and transmission ratios of 93Nb were measured by applying the time-of-flight method using the Accurate Neutron-Nucleus Reaction measurement Instrument of the Materials and Li...

A pilot study to quantify 55Fe in steel from a reactor vessel of a nuclear power plant by accelerator mass spectrometry (AMS) without any chemical sample preparation was validated by liquid scintillation counting (LSC) and AMS after radiochemical separation. AMS reaches an uncertainty Fe−1 level within less than 10 min measuring time. The background was Fe−1, presently limited by the short measurement time. The new instrumental AMS method for analysing 55Fe from neutron capture production is reasonable and fast compared to other analytical methods.

In this study, Monte Carlo codes, Geant4 and MCNP6, were used to characterize the fast neutron therapeutic beam produced at iThemba LABS in South Africa. Experimental and simulation results were compared using the latest generation of Silicon on Insulator (SOI) microdosimeters from the Centre for Medical Radiation Physics (CMRP). Geant4 and MCNP6 were able to successfully model the neutron gantry and simulate the expected neutron energy spectrum produced from the reaction by protons bombarding a 9Be target. The neutron beam was simulated in a water phantom and its characteristics recorded by the silicon microdosimeters; bare and covered by a 10B enriched boron carbide converter, at different positions. The microdosimetric quantities calculated using Geant4 and MCNP6 are in agreement with experimental measurements. The thermal neutron sensitivity and production of 10B capture products in the p+ boron-implanted dopant regions of the Bridge microdosimeter is investigated. The obtained results are useful for the future development of dedicated SOI microdosimeters for Boron Neutron Capture Therapy (BNCT). This paper provides a benchmark comparison of Geant4 and MCNP6 capabilities in the context of further applications of these codes for neutron microdosimetry.

Canada’s access to neutron beams for neutron scattering was significantly curtailed in 2018 with the closure of the National Research Universal (NRU) reactor in Chalk River, Ontario, Canada. New sources are needed for the long-term; otherwise, access will only become harder as the global supply shrinks. Compact Accelerator-based Neutron Sources (CANS) offer the possibility of an intense source of neutrons with a capital cost significantly lower than spallation sources. In this paper, we propose a CANS for Canada. The proposal is staged with the first stage offering a medium neutron flux, linear accelerator-based approach for neutron scattering that is also coupled with a boron neutron capture therapy (BNCT) station and a positron emission tomography (PET) isotope production station. The first stage will serve as a prototype for a second stage: a higher brightness, higher cost facility that could be viewed as a national centre for neutron applications.

The thermal neutron porosity is routinely acquired in almost every well. When combined with the density, gamma ray and resistivity logs, the basic petrophysical parameters of a reservoir are evaluated. The design of the thermal neutron tool is simple, but its interpretation is complex and affected by the formation constituents. The most challenging situation occurs when the formation contains elements with high absorption probability of the thermal neutrons. The existence of such elements changes the neutron transport parameters and results in a false increase in the measured porosity. The problem is reported by many users throughout the years. In 1993, higher thermal neutron porosity is reported due to the existence of an iron-rich mineral, Siderite, in the Nazzazat and Baharia formations in Egypt. Siderite and all iron-rich minerals have high thermal neutrons absorption probability. Recently, in 2018, high thermal neutron porosity in Unayzah field in Saudi Arabia is also reported due to the existence of few parts per million of gadolinium. Gadolinium is a rare element that has high probability of thermal neutron absorption. Currently, none of the existing commercial petrophysics software(s) have modules to correct the thermal neutron porosity for such effects. This represents a challenge to the petrophysicists to properly calculate the actual reservoir porosity. In this paper, the effects of the rare elements and other minerals with high thermal neutron absorption probability on the thermal neutron porosity are discussed, and a correction methodology is developed and tested. The methodology is based on integrating the tool design and the physics of the neutron transport to perform the correction. The details of the correction steps and the correction algorithm are included, tested and applied in two fields.

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.

The isomeric ratios in neutron capture reactions, induced by thermal, resonant and mixed thermal-resonant neutrons on 114Cd and 116Cd nuclei have been measured by the activation method using off-line gamma spectroscopy. The data analysis and necessary corrections as the self-absorption and summing cascade effects were made in gamma rays radioactivity measurements to improve the precision of the isomeric ratio determination. The results were discussed and compared with those in the existing literature, which can contribute as additional data to the nuclear data banks and used for theoretical calculations in nuclear reaction models as well as for nuclear applications.

Reliable data on 7Li(p,p'γ)7Li reaction cross section and photon yield from a thick lithium target are important for many applications including boron neutron capture therapy of malignant tumors. Experimental data on cross section differ greatly from one author to another; experimental data on photon yield are extremely scarce. Measurements of the reaction cross section and photon yield were carried out at the accelerator-based neutron source at the Budker Institute of Nuclear Physics (Novosibirsk, Russia) using a NaI and HPGe γ-ray spectrometers. The 7Li(p,p'γ)7Li reaction cross section and 478 keV photon yield from a thick lithium target at proton energies from 0.65 MeV to 2.225 MeV have been measured with high accuracy. The data obtained are presented in the form of tables.

Purpose Boron Neutron Capture Therapy (BNCT) is a form of hadrontherapy based on the selective damage caused by the products of neutron capture in 10B to tumour cells. BNCT dosimetry strongly depends on the parameters of the dose calculation models derived from radiobiological experiments. This works aims at determining an adequate dosimetry for in-vitro experiments involving irradiation of monolayer-cultured cells with photons and BNCT and assessing its impact on clinical settings. M&M Dose calculations for rat osteosarcoma UMR-106 and human metastatic melanoma Mel-J cell survival experiments were performed using MCNP, transporting uncharged particles for KERMA determinations, and secondary particles (electrons, protons, 14C, 4He and 7Li) to compute absorbed dose in cultures. Dose-survival curves were modified according to the dose correction factors determined from computational studies. New radiobiological parameters of the photon isoeffective dose models for osteosarcoma and metastatic melanoma tumours were obtained. Dosimetry implications considering cutaneous melanoma patients treated in Argentina with BNCT were assessed and discussed. Results KERMA values for the monolayer-cultured cells overestimate absorbed doses of radiation components of interest in BNCT. Detailed dose calculations for the osteosarcoma irradiation increased the relative biological effectiveness factor RBE1% of the neutron component in more than 30%. The analysis based on melanoma cases reveals that the use of survival curves based on KERMA leads to an underestimation of the tumour doses delivered to patients. Conclusions Considering detailed dose calculation for in-vitro experiments significantly impact on the prediction of the tumor control in patients. Therefore, proposed methods are clinically relevant.

Understanding the formation, structure and evolution of the Milky Way is one of the main open questions in modern astronomy. The Galaxy is made up of chemical elements originated from Big Bang to stellar nucleosynthesis. Among them, there are some that stand out for their special nucleosynthesis: they are the so-called neutron-capture elements. Although they are the majority among the elements, they only represent a small amount of the global composition of the Universe. However, it is in this small percentage where the keys of the evolution of the Galaxy are found. Among these keys are the chemical abundances. Based on 506 stars of the Solar neighbourhood observed at the Molėtai Astronomical Observatory and 62 open clusters from the Gaia ESO Survey, this thesis tried to answer some of the main questions in this field. Through its 6 chapters, the thesis deals with extremely important topics such as the temporal and spatial abundance gradients of neutron-capture elements in the Galactic thin and thick discs, as well as their application to calculate the ages of stars. Among its most significant results, the thesis provides new insights on s-processes indicators and the astrophysical origin of the r-process. For the first time a comprehensive analysis of the use of chemical clocks in the thick disc is made and the radial variations of calibrated chemical clocks are investigated in depth using a reliable sample of open clusters, indeed the largest used for this purpose.

Neutron capture cross-section measurements for 237Np have been conducted with the Accurate Neutron Nucleus Reaction Measurement Instrument (ANNRI) at the Materials and Life Science Facility (MLF) ...

We have been developing a new-generation accelerator named “Skeleton Cyclotron”, which is an ultra-compact and high-intensity accelerator for the stable production of 211At used for targeted alpha-particle therapy. Coil system for Skeleton Cyclotron consists of circular and noncircular high-temperature superconducting magnets and it can cause the output energy to vary without being bounded by the non-linearity of the iron core. This will enable not only the production of 211At, but also the production of radioisotopes for positron emission tomography and neutron irradiation for boron neutron capture therapy with a single accelerator. A high-precision magnetic field distribution and a high temporal stability are required for stable beam acceleration, but the screening current, which is induced in the REBa2Cu3Ox (REBCO, RE: Rare Earth) coated conductor, generates an irregular magnetic field and deteriorates the quality of the magnetic field spatially and temporally. In this study, we numerically evaluated the temporal stability and reproducibility of the screening current-induced field considering the variable energy operation for the Ultra-Baby Skeleton Cyclotron, which is a small demonstration model of Skeleton Cyclotron.

PURPOSE The stochastic microdosimetric kinetic (SMK) model is one of the most sophisticated and precise models used in the estimation of the relative biological effectiveness of carbon-ion radiotherapy (CRT) and boron neutron capture therapy (BNCT). However, because of its complicated and time-consuming calculation procedures, it is nearly impractical to directly incorporate this model into a radiation treatment-planning system. MATERIALS AND METHODS Through the introduction of Taylor expansion (TE) or fast Fourier transform (FFT), we developed two simplified SMK models and implemented them into the Particle and Heavy Ion Transport code System (PHITS). To verify the implementation, we calculated the photon isoeffective doses in a cylindrical phantom placed in the radiation fields of passive CRT and accelerator-based BNCT. RESULTS AND DISCUSSION Our calculation suggested that both TE-based and FFT-based SMK models can reproduce the data obtained from the original SMK model very well for absorbed doses approximately below 5 Gy, whereas the TE-based SMK model overestimates the original data at higher doses. In terms of computational efficiency, the TE-based SMK model is much faster than the FFT-based SMK model. CONCLUSION This study enables the instantaneous calculation of the photo isoeffective dose for CRT and BNCT, considering their cellular-scale dose heterogeneities. Treatment-planning systems that use the improved PHITS as a dose-calculation engine are under development.

The Self Powered Neutrons Detectors (SPND) have the advantage of not requiring a high voltage power supply for their operation and are small in size, enhancing the interest of these detectors in medicine. In this context, we have developed a thermal neutron detection system based on SPND. This detector was placed in the thermal channel of our nuclear research reactor; where the values of the current for each detector have been recorded as a function of time, with a chain in a current mode where electrometers without HV were used. We performed the real-time measurement of neutron flux during boron neutron capture therapy or boron neutron therapy, the different materials constituting the SPND detectors have been carefully chosen for this application. These detectors were tested at a power of four MW corresponding to a neutron flux of 109 n cm−2 s−1. The usefulness of 103Rh-SPND is for online measurement of thermal neutron flux on BNCT patients has been demonstrated based on an appropriate calibration of the thermal neutron spectrum.

We study the enrichment of the interstellar medium with rapid neutron capture (r-process) elements produced in binary neutron star (BNS) mergers. We use a semi-analytic model to describe galactic evolution, with merger rates and time delay distributions of BNS mergers consistent with the latest population synthesis models. In order to study the dispersion of the relative abundances of r-process elements and iron, we applied a turbulent mixing scheme, where the freshly synthesized elements are gradually dispersed in the interstellar medium. We show that within our model the abundances observed in Milky-Way stars, in particular the scatter at low metallicities, can be entirely explained by BNS mergers. Our results suggest that binary neutron star mergers could be the dominant source of r-process elements in the Galaxy., Comment: 16 pages, 19 figures. Modified to match the version accepted by MNRAS (some changes to main results and conclusions, new section added)

The nMCP (Neutron sensitive microchannel plate) combined with advanced readout electronics is widely used in energy selective neutron imaging because of its good spatial and timing resolution. Neutron detection efficiency is a crucial parameter for the nMCP. In this paper, a mathematical model based on the oblique cylindrical channel and elliptical pore was established to calculate the neutron absorption probability, the escape probability of charged particles and overall detection efficiency of nMCP and analyze the effects of neutron incident position, pore diameter, wall thickness and bias angle. It was shown that when the doping concentration of the nMCP was 10 mole%, the thickness of nMCP was 0.6 mm, the detection efficiency could reach maximum value, about 24% for thermal neutrons if the pore diameter was 6μm, the wall thickness was 2μm and the bias angle was 3 or 6 degrees. The calculated results are of great significance for evaluating the detection efficiency of the nMCP. In a subsequent companion paper, the mathematical model would be extended to the case of the spatial resolution and detection efficiency optimization of the coating nMCP.

A linear fast neutron beam chopper has been deployed at a prompt gamma activation analysis instrument, enabling in-beam activation analysis of short-lived ( $T_{\mathrm {1/2}} > \sim 10$ ms) neutron capture products without the need for sample transfer. This article describes the design and operation of the chopper system and measurement results from test samples containing Se-77 ( $T_{\mathrm {1/2}} =17.36$ s) and Yb-175m ( $T_{\mathrm {1/2}} =0.0682$ s). The chopper timing was optimized for a specific isotope in each measurement, demonstrating an advantage of the programmable system. The linear response of the total gated counts in the decay phase to the incremented Se and Yb masses demonstrates the suitability of the system for quantitative analysis. Theoretical calculations based on the experimental parameters were in good agreement with the observed results. An example that illustrates the separation of overlapping energy peaks by the use of time-stamped list mode data acquisition is shown in a sample containing Yb-175m and Dy-165 ( $T_{\mathrm {1/2}} =75.4$ s).

Introduction. Boron neutron capture therapy (bnct) is a promising method for treating tumors, in particular, infiltrative malignant tumors, due to the selective destruction of tumor cells without damaging the surrounding normal tissues. This type of therapy is based on nuclear reaction of neutron capture by stable 10b isotope. For the successful implementation of bnct, boron delivery drugs that must be selectively accumulated in malignant cells in a sufficient amount, and a neutron source with the energy required for the neutron capture reaction are needed. At the budker institute of nuclear physics, the accelerator-based neutron source was designed with flux parameters allowing studies on bnct to be conducted.Objective: to assess the effect of bnct on tumor and normal cell lines using borphenylalanine (bpa), borcaptate (bsh) and liposomal borcaptat as boron delivery drugs.Materials and methods. Human cell cultures: glioblastoma (u87), colorectal human adenocarcinoma (sw-620), human melanoma (sk-mel28) and primary embryonic cell lines were irradiated with a neutron flux at the presence of bpa, bsh and liposomal bsh with a concentration of 10b 40 μg/ml. The short-term cytotoxic effect of irradiation was evaluated using trypan blue. Cell survival 96 hours after irradiation was determined using mtt test, and survival fraction was evaluated using the clonogenic test.Results. Early cytotoxic effects of irradiation were not observed for all 4 cell lines. According to mtt and clonogenic tests, the most pronounced effect of bnct was noticed for sw-620 and u87 lines, regardless of boron delivery drug used. For sk-mel28 line, the best effect was achieved after irradiation with liposomal borocaptate. For the primary transplanted embryonic line, high toxicity was revealed when bnct was performed with borphenylalanine and borcaptate.Conclusion. The data obtained indicate that the accelerator-based bnct using boron delivery drugs, such as borphenylalanine, borcaptate and liposomal borcaptat, has a positive effect on tumor lines of glioblastoma, colorectal adenocarcinoma and melanoma.

The aim of this study was to optimize the target, moderator, and collimator (TMC) in a neutron beam generator for the accelerator-based BNCT (A-BNCT) system. The optimization employed the Monte Carlo Neutron and Photon (MCNP) simulation. The optimal geometry for the target was decided as the one with the highest neutron flux among nominates, which were called as angled, rib, and tube in this study. The moderator was optimized in terms of consisting material to produce appropriate neutron energy distribution for the treatment. The optimization of the collimator, which wrapped around the target, was carried out by deciding the material to effectively prevent the leakage radiations. As results, characteristic of the neutron beam from the optimized TMC was compared to the recommendation by the International Atomic Energy Agent (IAEA). The tube type target showed the highest neutron flux among nominates. The optimal material for the moderator and collimator were combination of Fluental (Al203+AlF3) with 60Ni filter and lead, respectively. The optimized TMC satisfied the IAEA recommendations such as the minimum production rate of epithermal neutrons from thermal neutrons: that was 2.5 times higher. The results can be used as source terms for shielding designs of treatment rooms.

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.

There has recently been intense research on boron nitride (BN)-based materials for radiation shielding, including shielding neutron radiations from nuclear reactors. However, advanced materials should display more than just radiation-shielding functions to be considered to be an outstanding candidate for applications in space; consequently, multifunctional BN-based materials with high hydrogen intake have been found to provide sufficient shielding properties. This article reviews the BN nanostructures and BN-based materials for space radiation-shielding application. It begins with a brief introduction of BN structures and their physical properties and followed by an emphasis on radiation-shielding properties of BN as related to neutron capture and hydrogen storage. In addition, development of BN-based materials for a space radiation-shielding application is discussed.

Natural plutonium from supernovaeThe rapid neutron capture process (r-process) produces many of the heavy chemical elements, but the astrophysical settings where it occurs remain unclear. Leading candidates are neutron star mergers and some types of supernovae. Wallneret al.analyzed the plutonium content of a deep-sea crust sample, identifying a few dozen atoms of the r-process isotope plutonium- 244 that were delivered to Earth within the past few million years. There was a simultaneous signal of iron-60, which is known to be produced in supernovae. Comparing the ratios of these isotopes constrains the relative contributions of supernovae and neutron star mergers to r-process nucleosynthesis.Science, this issue p.742

The neutron reflectometry method is used to measure the spatial profile of the potential of interaction between neutrons and a medium. At the interface between media, the interaction potential is the sum of potentials of individual isotopes. Neutrons and gamma radiation emitted by the atomic nuclei of an element are recorded to determine the potential of interaction between neutrons and individual isotopes. Channels for the recording of gamma quanta and spin-flipped neutrons are created using a REMUR spectrometer. Channel testing results are given, model calculations of the neutron absorption coefficients are carried out, and prospects related to the recording of gamma quanta and polarized neutrons are discussed.

The research on new radiopharmaceuticals for therapy of cancer is evolving rapidly. Thanks to novel technologies and new selective and less toxic compounds, we move towards personalized molecular medicine. The neutron capture radiation therapy (NCT) can be potentially much safer and can offer a better spatial and temporal control than the radioisotope therapy. Still, there are not many options in NCT: the 10B isotope has been almost exclusively used for decades, and only recently, 157Gd has attracted some interest. Here, we want to draw attention to a new nuclide, 7Be, recently suggested for the NCT, and discuss perspective of Be2+ confinement in aqueous solutions and targeted delivery to cancerous tissues.

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.

Abstract Refractory ‘stuffed’ pyrochlores such as Gd2TiO5 are of interest for nuclear applications, including as matrices for actinide disposition and as neutron absorbers in control rods. Here, we report the results of a preliminary comparative investigation of the synthesis of Gd2TiO5 by molten salt and conventional solid-state synthesis. We show that synthesis of Gd2TiO5 proceeds from the pyrochlore phase Gd2Ti2O7 which is first formed as a kinetic product. Molten salt synthesis afforded single phase Gd2TiO5 at 1300 °C in 2 h, via a template growth mechanism, and is effective for the synthesis of these refractory materials. This work demonstrates molten salt mediated synthesis of ‘stuffed’ pyrochlore for the first time. Graphic abstract

The use of the chromatographic method to extract 99Tcm from the low specific activity 99Mo solution is an effective way to obtain medical isotope 99Tcm that meeting clinical requirements. It has been reported that the separation technologies must use strong acid or strong alkaline solutions, which require additional treatments. Exploring the method of separating technetium from neutral molybdenum solution can simplify the treatment process. A kind of activated carbon fiber was used to extract technetium from neutral low specific activity 99Mo (＜0.2 Ci/g Mo) in 1 mol/L NaCl solution. Neutral eluent was used to desorb technetium throughout the process, and the separation of molybdenum and technetium was realized. The recoveries of 99Tcm were more than 90%. The obtained sodium pertechnetate ［99Tcm］ injection met the requirements of the Chinese Pharmacopoeia without pH adjustation, and the labeling rate of MIBI and MDP for met the Chinese Pharmacopoeia claim. The results show that the extraction of 99Tcm from 99Mo with low specific activity (

The compound 4-dihydroxyboryl-l-phenylalanine (BPA) has found use in clinical trials of boron neutron capture therapy (BNCT). Here, we have examined the interaction with DNA of an amide-blocked BPA derivative of hexa-l-arginine (Ac-BPA-Arg6-NH2). Physical and spectroscopic assays show that this peptide binds to and condenses DNA. The resulting condensates are highly resistant to the effects of nuclease incubation (68-fold) and gamma (38-fold) irradiation. Radioprotection was modeled by Monte Carlo track structure simulations of DNA single strand breaks (SSBs) with TOPAS-nBio. The differences between experimental and simulated SSB yields for uncondensed and condensed DNAs were ca. 2 and 18%, respectively. These observations indicate that the combination of a plasmid DNA target, the BPA-containing peptide, and track structure simulation provides a powerful approach to characterize DNA damage by the high-LET radiation associated with neutron capture on boron.

Fundamentals of boron neutron capture therapy of malignant tumors are outlined, and the current status of the development of epithermal-neutron sources on the basis of charged-particle accelerators is surveyed. Attention is additionally given to a neutron source based on a charged-particle accelerator belonging to a new type—a tandem accelerator with vacuum insulation and a lithium neutron-producing target.

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Analysis Activation) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron–gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

Background: Due to the noninvasive nature of boron neutron capture therapy (BNCT), it is considered a promising cancer treatment method. Aim: To investigate whether polyvinyl alcohol/boric acid crosslinked nanoparticles (PVA/BA NPs) are an efficient delivery system for BNCT. Materials & methods: PVA/BA NPs were synthesized and cocultured with brain and oral cancers cells for BNCT. Results: PVA/BA NPs had a boron-loading capacity of 7.83 ± 1.75 w/w%. They accumulated in brain and oral cancers cells at least threefold more than in fibroblasts and macrophages. The IC50 values of the brain and oral cancers cells were at least ninefold and sixfold lower than those of fibroblasts and macrophages, respectively. Conclusion: Theoretically, PVA/BA NPs target brain and oral cancers cells and could offer improved therapeutic outcomes of BNCT.

The photon and neutron absorbing capacity of titanate-doped borate glasses: 65B(2)O(3)-30Li(2)O-5Al(2)O(3)-xTiO(2): x = 0-30 mol% coded as G1-G7 were investigated via WinXCOM and EXABCal computer codes. Mass (mu(m)) and linear (LAC) attenuation coefficients, mean-free path (MFP), half-value thickness (HVT), buildup factors (EABUF and EBUF), and the relative fast neutron absorbing efficacy were investigated. The maximum value of mu(m) at 15 keV was 1.68, 2.976, 4.119, 5.134, 6.042, 6.856, and 7.593 cm(2)/g for G1, G2, G3, G4, G5, G6, and G7, respectively. For energies below 0.1 MeV and above 10 meV, mu(m) values vary according to the order (G1)(mu m) < (G2)(mu m) < (G3)(mu m) < (G4)(mu m) < (G5)(mu m) < (G6)(mu m) < (G7)(mu m). Values of LAC followed the order: (G7)(LAC) > (G6)(LAC) > (G5)(LAC) > (G4)(LAC) > (G3)(LAC) > (G2)(LAC) > (G1)(LAC). The trend of the MFP increases among the glasses follows a reverse order as that of LAC: (G1)(MFP) > (G2)(MFP) > (G3)(MFP) > (G4)(MFP) > (G5)(MFP) > (G6)(MFP) > (G7)(MFP). The HVT of the glasses follow the order: (G1)(HVT) > (G2)(HVT) > (G3)(HVT) > (G4)(HVT) > (G5)(HVT) > (G6)(HVT) > (G7)(HVT). EABUF and EBUF increase in the order (G1)((EABUF, EBUF)) > (G2)((EABUF, EBUF)) > (G3)((EABUF, EBUF)) > (G4)((EABUF, EBUF)) > (G5)((EABUF, EBUF)) > (G6)((EABUF, EBUF)) > (G7)((EABUF, EBUF)). The investigated glasses (G1-G7) are better fast neutron absorbers compared to ordinary concrete and water. Results revealed that higher titanate doping concentrations produced better photon and fast neutron shielding capacities. Therefore, the investigated glasses showed superior shielding efficacy when compared with some traditional shielding materials; consequently, the G1-G7 glasses are recommended for use as ionizing radiation shields.

Liver cancer was the third leading cause of death from cancer in 2020 with 830,180 deaths worldwide. Radiotherapy is a common treatment method for liver cancer. Technological advances presented proton therapy and boron neutron capture therapy (BNCT) as alternatives with a lower dose on healthy organs. The objective of this research is to get a good dose distribution with higher tumor dose and lower healthy organ dose in proton therapy. A comparison with BNCT is done to get a better understanding of how both methods deliver the dose to treat the cancer while minimizing healthy organ doses. The research simulated proton therapy for cancer liver with Particle and Heavy Ions Transport Code System (PHITS), and a literature review for BNCT. The effectiveness of both methods were compared by tumor dose and liver dose. The optimal tumor dose for proton therapy is 86.01 Gy (W) with 0.67 Gy (W) liver dose. Proton therapy can replace conventional radiotherapy for tumors with complex shapes in dose delivery by utilizing its dose profile, while BNCT can give better tumor control on patients previously treated with conventional radiotherapy.

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.

Boron–neutron capture therapy (BNCT) is a cancer treatment method that exploits the high neutron reactivity of boron. Monitoring the prompt gamma rays (PGs) produced during neutron irradiation is essential for ensuring the accuracy and safety of BNCT. We investigate the imaging of PGs produced by the boron–neutron capture reaction through Monte Carlo simulations of a gamma camera with a SrI2 scintillator and parallel-hole collimator. GAGG scintillator is also used for a comparison. The simulations allow the shapes of the energy spectra, which exhibit a peak at 478 keV, to be determined along with the PG images from a boron–water phantom. It is found that increasing the size of the water phantom results in a greater number of image counts and lower contrast. Additionally, a higher septal penetration ratio results in poorer image quality, and a SrI2 scintillator results in higher image contrast. Thus, we can simulate the BNCT process and obtain an energy spectrum with a reasonable shape, as well as suitable PG images. Both GAGG and SrI2 crystals are suitable for PG imaging during BNCT. However, for higher imaging quality, SrI2 and a collimator with a lower septal penetration ratio should be utilized.