In-Operando Lithium-Ion Transport Tracking in an All-Solid-State Battery.

An all-solid-state battery is a secondary battery that is charged and discharged by the transport of lithium ions between positive and negative electrodes. To fully realize the significant benefits of this battery technology, for example, higher energy densities, faster charging times, and safer operation, it is essential to understand how lithium ions are transported and distributed in the battery during operation. However, as the third lightest element, methods for quantitatively analyzing lithium during operation of an all-solid-state device are limited such that real-time tracking of lithium transport has not yet been demonstrated. Here, the authors report that the transport of lithium ions in an all-solid-state battery is quantitatively tracked in near real time by utilizing a high-intensity thermal neutron source and lithium-6 as a tracer in a thermal neutron-induced nuclear reaction. Furthermore, the authors show that the lithium-ion migration mechanism and pathway through the solid electrolyte can be determined by in-operando tracking. From these results, the authors suggest that the development of all-solid-state batteries has entered a phase where further advances can be carried out while understanding the transport of lithium ions in the batteries.

Characterization of Thermoluminescent Dosimeters for Neutron Dosimetry at High Altitudes.

Neutrons constitute a significant component of the secondary cosmic rays and are one of the most important contributors to natural cosmic ray radiation background dose. The study of the cosmic ray neutrons' contribution to the dose equivalent received by humans is an interesting and challenging task for the scientific community. In addition, international regulations demand assessing the biological risk due to radiation exposure for both workers and the general population. Because the dose rate due to cosmic radiation increases significantly with altitude, the objective of this work was to characterize the thermoluminescent dosimeter (TLDs) from the perspective of exposing them at high altitudes for longtime neutron dose monitoring. The pair of TLD-700 and TLD-600 is amply used to obtain the information on gamma and neutron dose in mixed neutron-gamma fields due to the present difference in 6Li isotope concentration. A thermoluminescence dosimeter system based on pair of TLD-600/700 was characterized to enable it for neutron dosimetry in the thermal energy range. The system was calibrated in terms of neutron ambient dose equivalent in an experimental setup using a 241Am-B radionuclide neutron source coated by a moderator material, polyethylene, creating a thermalized neutron field. Afterward, the pair of TLD-600/700 was exposed at the CERN-EU High-Energy Reference Field (CERF) facility in Geneva, which delivers a neutron field with a spectrum similar to that of secondary cosmic rays. The dosimetric system provided a dose value comparable with the calculated one demonstrating a good performance for neutron dosimetry.

High-Accuracy Relative Biological Effectiveness Values Following Low-Dose Thermal Neutron Exposures Support Bimodal Quality Factor Response with Neutron Energy.

Theoretical evaluations indicate the radiation weighting factor for thermal neutrons differs from the current International Commission on Radiological Protection (ICRP) recommended value of 2.5, which has radiation protection implications for high-energy radiotherapy, inside spacecraft, on the lunar or Martian surface, and in nuclear reactor workplaces. We examined the relative biological effectiveness (RBE) of DNA damage generated by thermal neutrons compared to gamma radiation. Whole blood was irradiated by 64 meV thermal neutrons from the National Research Universal reactor. DNA damage and erroneous DNA double-strand break repair was evaluated by dicentric chromosome assay (DCA) and cytokinesis-block micronucleus (CBMN) assay with low doses ranging 6-85 mGy. Linear dose responses were observed. Significant DNA aberration clustering was found indicative of high ionizing density radiation. When the dose contribution of both the 14N(n,p)14C and 1H(n,γ)2H capture reactions were considered, the DCA and the CBMN assays generated similar maximum RBE values of 11.3 ± 1.6 and 9.0 ± 1.1, respectively. Consequently, thermal neutron RBE is approximately four times higher than the current ICRP radiation weighting factor value of 2.5. This lends support to bimodal peaks in the quality factor for RBE neutron energy response, underlining the importance of radiological protection against thermal neutron exposures.

Possible External Factors Determining Ultradian (4-20-min) Rhythms of Body Temperature.

The study examined the effect of passive magnetic shielding on the magnitude of rhythmic oscillations of body temperature (BT) with 4-20 min periods in mice and their correlation with similar oscillations in unshielded control group. A magnetic permalloy screen that 35-fold attenuates the total geomagnetic field and decreased the undulations of magnetic field with the periods of few minutes by 5 times, produced no effect on the mean amplitude of BT oscillations within the same period range, their spectral power, and the cross-spectral density of examined rhythms in comparison with the control (unshielded) mice. Thus, either the mice possess a very sensitive magnetic sensory system or there exists an external non-magnetic factor affecting rhythmicity of BT. The study advanced intensity of thermal neutron radiation near the Earth's surface known to reflect the flow of accelerated particles generated by the secondary cosmic rays as the external factor, which strongly correlates with BT rhythms revealed by cross-spectrum analysis.

Determination of the thermal and epithermal neutron sensitivities of an LBO chamber.

An LBO (Li2B4O7) walled ionization chamber was designed to monitor the epithermal neutron fluence in boron neutron capture therapy clinical irradiation. The thermal and epithermal neutron sensitivities of the device were evaluated using accelerator neutrons from the 9Be(d, n) reaction at a deuteron energy of 4 MeV (4 MeV d-Be neutrons). The response of the chamber in terms of the electric charge induced in the LBO chamber was compared with the thermal and epithermal neutron fluences measured using the gold-foil activation method. The thermal and epithermal neutron sensitivities obtained were expressed in units of pC cm2, i.e., from the chamber response divided by neutron fluence (cm-2). The measured LBO chamber sensitivities were 2.23 × 10-7 ± 0.34 × 10-7 (pC cm2) for thermal neutrons and 2.00 × 10-5 ± 0.12 × 10-5 (pC cm2) for epithermal neutrons. This shows that the LBO chamber is sufficiently sensitive to epithermal neutrons to be useful for epithermal neutron monitoring in BNCT irradiation.

A re-activation model for 169Yb intensity modulated brachytherapy sources accounting for spatiotemporal isotopic composition.

BACKGROUND: Intensity modulated brachytherapy based on partially shielded intracavitary and interstitial applicators is possible with a cost-effective 169Yb production method. 169Yb is a traditionally expensive isotope suitable for this purpose, with an average γ-ray energy of 93 keV. Re-activating a single 169Yb source multiple times in a nuclear reactor between clinical uses was shown to theoretically reduce cost by approximately 75% relative to conventional single-activation sources. With re-activation, substantial spatiotemporal variation in isotopic source composition is expected between activations via 168Yb burnup and 169Yb decay, resulting in time dependent neutron transmission, precursor usage, and reactor time needed per re-activation. PURPOSE: To introduce a generalized model of radioactive source production that accounts for spatiotemporal variation in isotopic source composition to improve the efficiency estimate of the 169Yb production process, with and without re-activation. METHODS AND MATERIALS: A time-dependent thermal neutron transport, isotope transmutation, and decay model was developed. Thermal neutron flux within partitioned sub-volumes of a cylindrical active source was calculated by raytracing through the spatiotemporal dependent isotopic composition throughout the source, accounting for thermal neutron attenuation along each ray. The model was benchmarked, generalized, and applied to a variety of active source dimensions with radii ranging from 0.4 to 1.0 mm, lengths from 2.5 to 10.5 mm, and volumes from 0.31 to 7.85 mm3, at thermal neutron fluxes from 1 × 1014 to 1 × 1015 n cm-2 s-1. The 168Yb-Yb2O3 density was 8.5 g cm-3 with 82% 168Yb-enrichment. As an example, a reference re-activatable 169Yb active source (RRS) constructed of 82%-enriched 168Yb-Yb2O3 precursor was modeled, with 0.6 mm diameter, 10.5 mm length, 3 mm3 volume, 8.5 g cm-3 density, and a thermal neutron activation flux of 4 × 1014 neutrons cm-2 s-1. RESULTS: The average clinical 169Yb activity for a 0.99 versus 0.31 mm3 source dropped from 20.1 to 7.5 Ci for a 4 × 1014 n cm-2 s-1 activation flux and from 20.9 to 8.7 Ci for a 1 × 1015 n cm-2 s-1 activation flux. For thermal neutron fluxes ≥2 × 1014 n cm-2 s-1, total precursor and reactor time per clinic-year were maximized at a source volume of 0.99 mm3 and reached a near minimum at 3 mm3. When the spatiotemporal isotopic composition effect was accounted for, average thermal neutron transmission increased over RRS lifetime from 23.6% to 55.9%. A 28% reduction (42.5 days to 30.6 days) in the reactor time needed per clinic-year for the RRS is predicted relative to a model that does not account for spatiotemporal isotopic composition effects. CONCLUSIONS: Accounting for spatiotemporal isotopic composition effects within the RRS results in a 28% reduction in the reactor time per clinic-year relative to the case in which such changes are not accounted for. Smaller volume sources had a disadvantage in that average clinical 169Yb activity decreased substantially below 20 Ci for source volumes under 1 mm3. Increasing source volume above 3 mm3 adds little value in precursor and reactor time savings and has a geometric disadvantage.

Biomarker dosimetry of acute low level of thermal neutrons and radiation adaptive response effect on rats.

In this paper, we demonstrated the biological effects of acute low-dose neutrons on the whole body of rats and investigated the impact of that level of neutron dose to induce an in vivo radio-adaptive response. To understand the radio-adaptive response, the examined animals were exposed to acute neutron radiation doses of 5 and 10 mSv, followed by a 50 mSv challenge dose after 14 days. After irradiation, all groups receiving single and double doses were kept in cages for one day before sampling. The electron paramagnetic resonance (EPR) method was used to estimate the radiation-induced radicals in the blood, and some hematological parameters and lipid peroxidation (MDA) were determined. A comet assay was performed beside some of the antioxidant enzymes [catalase enzyme (CAT), superoxide dismutase (SOD), and glutathione (GSH)]. Seven groups of adult male rats were classified according to their dose of neutron exposure. Measurements of all studied markers are taken one week after harvesting, except for hematological markers, within 2 h. The results indicated lower production of antioxidant enzymes (CAT by 1.18-5.83%, SOD by 1.47-17.8%, and GSH by 11.3-82.1%). Additionally, there was an increase in red cell distribution width (RDW) (from 4.61 to 25.19%) and in comet assay parameters such as Tail Length, (from 6.16 to 10.81 µm), Tail Moment, (from 1.17 to 2.46 µm), and percentage of DNA in tail length (DNA%) (from 9.58 to 17.32%) in all groups exposed to acute doses of radiation ranging from 5 to 50 mSv, respectively. This emphasizes the ascending harmful effect with the increased acute thermal neutron doses. The values of the introduced factor of radio adaptive response for all markers under study reveal that the lower priming dose promotes a higher adaptation response and vice versa. Ultimately, the results indicate significant variations in DNA%, SOD enzyme levels, EPR intensity, total Hb concentration, and RDWs, suggesting their potential use as biomarkers for acute thermal neutron dosimetry. Further research is necessary to validate these measurements as biodosimetry for radiation exposure, including investigations involving the response impact of RAR with varied challenge doses and post-irradiation behavior.

A CMOS-MEMS Pixel Sensor for Thermal Neutron Imaging.

A monolithic pixel sensor with high spatial granularity (35 × 40 µm2) is presented, aiming at thermal neutron detection and imaging. The device is made using the CMOS SOIPIX technology, with Deep Reactive-Ion Etching post-processing on the backside to obtain high aspect-ratio cavities that will be filled with neutron converters. This is the first monolithic 3D sensor ever reported. Owing to the microstructured backside, a neutron detection efficiency up to 30% can be achieved with a 10B converter, as estimated by the Geant4 simulations. Each pixel includes circuitry that allows a large dynamic range and energy discrimination and charge-sharing information between neighboring pixels, with a power dissipation of 10 µW per pixel at 1.8 V power supply. The initial results from the experimental characterization of a first test-chip prototype (array of 25 × 25 pixels) in the laboratory are also reported, dealing with functional tests using alpha particles with energy compatible with the reaction products of neutrons with the converter materials, which validate the device design.

Large Scale BN-perovskite Nanocomposite Aerogel Scintillator for Thermal Neutron Detection.

State-of-the-art thermal neutron scintillation detectors rely on rare isotopes for neutron capture, lack stability and scalability of solid-state scintillation devices, and poorly discriminate between the neutron and gamma rays. The boron nitride (BN)-CsPbBr3 perovskite nanocomposite aerogel scintillator enables discriminative detection of thermal neutrons, features the largest known size (9 cm across), the lowest density (0.17 g cm-3 ) among the existing scintillation materials, high BN (50%) perovskite (1%) contents, high optical transparency (85%), and excellent radiation stability. The new detection mechanism relies on thermal neutron capture by 10 B and effective energy transfer from the charged particles to visible-range scintillation photons between the densely packed BN and CsPbBr3 nanocrystals. Low density minimizes the gamma ray response. The neutrons and gamma rays are discriminated by complete decoupling of the respective single pulses in time and intensity. These outcomes open new avenues for neutron detection in resource exploration, clean energy, environmental, aerospace, and homeland security applications.

Selective detection of fast and thermal neutrons in mixed-radiation fields using tungsten-silica and gold-iodine-silica nanoparticles and their boron-loaded aqueous dispersions.

Tungsten-silica and gold-iodine-silica nanoparticles and their boron-loaded aqueous dispersions were used to selectively detect fast and thermal neutrons in mixed-radiation fields generated by a cyclotron on the order of mSv at a neutron flux of 1.0 ×106(neutron/sec∙cm2). The photo-image intensity, fluorescence spectra, absorption spectra, and XRD of their aqueous dispersions were measured immediately and eighteen days after irradiation. The immediate measurements of photo-image intensity and fluorescence spectral area ratios for gold-iodine-silica nanoparticle aqueous dispersions indicated the dose dependence of photo-image intensity and fluorescence spectral area ratios. Measurements of the relative fluorescence and absorption spectral areas of gold-iodine-silica nanoparticle aqueous dispersions 18 days after irradiation also showed similar dose dependences. The precipitates of gold-iodine-silica nanoparticles showed a linear relationship between the XRD peak ratio and the dose with a correlation coefficient of 0.9. The photo-image intensities, fluorescence spectral area, absorption spectral area, and XRD peak ratios were found to be affected by fast and thermal neutrons. Simple methods of fluorescence, absorption, and XRD measurements are proposed for the selective detection of fast and thermal neutrons in mixed-radiation fields.

Characterization of thermal neutron distribution of an Am-Be neutron source setup by CdZnTe detector.

Owing to the developments in prompt gamma neutron activation analysis (PGNAA) and prompt gamma ray activation imaging (PGAI), it is necessary to develop an online thermal neutron distribution measurement method. The CdZnTe detector is considered as an alternative thermal neutron detector because of its high thermal neutron capture cross section. In this study, the thermal neutron field of an 241Am-Be neutron source was determined by CdZnTe detector. The intrinsic neutron detection efficiency of CdZnTe detector was calculated by using indium foil activation, and the value was 3.65%. Then the characteristics of neutron source were conducted with the calibrated CdZnTe detector. The thermal neutron fluxes in front of beam port were measured at several distances ranging from 0 cm up to 28 cm. Thermal neutron field at distances of 1 cm and 5 cm were also measured. The experimental data were then compared with Monte Carlo simulation. The results demonstrated that the simulated data agree well with experimental measurements.

Development of optimization method for uniform dose distribution on superficial tumor in an accelerator-based boron neutron capture therapy system.

To treat superficial tumors using accelerator-based boron neutron capture therapy (ABBNCT), a technique was investigated, based on which, a single-neutron modulator was placed inside a collimator and was irradiated with thermal neutrons. In large tumors, the dose was reduced at their edges. The objective was to generate a uniform and therapeutic intensity dose distribution. In this study, we developed a method for optimizing the shape of the intensity modulator and irradiation time ratio to generate a uniform dose distribution to treat superficial tumors of various shapes. A computational tool was developed, which performed Monte Carlo simulations using 424 different source combinations. We determined the shape of the intensity modulator with the highest minimum tumor dose. The homogeneity index (HI), which evaluates uniformity, was also derived. To evaluate the efficacy of this method, the dose distribution of a tumor with a diameter of 100 mm and thickness of 10 mm was evaluated. Furthermore, irradiation experiments were conducted using an ABBNCT system. The thermal neutron flux distribution outcomes that have considerable impacts on the tumor's dose confirmed a good agreement between experiments and calculations. Moreover, the minimum tumor dose and HI improved by 20 and 36%, respectively, compared with the irradiation case wherein a single-neutron modulator was used. The proposed method improves the minimum tumor volume and uniformity. The results demonstrate the method's efficacy in ABBNCT for the treatment of superficial tumors.

Double-Focusing Thermal Triple-Axis Spectrometer at the NCNR.

The new thermal triple-axis spectrometer at the NIST Center for Neutron Research (NCNR) is located at the BT-7 beam port. The 165 mm diameter reactor beam is equipped with a selection of Söller collimators, beam-limiters, and a pyrolytic graphite (PG) filter to tailor the beam for the dual 20×20 cm(2) double-focusing monochromator system that provides monochromatic fluxes exceeding 10(8) n/cm(2)/s onto the sample. The two monochromators installed are PG(002) and Cu(220), which provide incident energies from 5 meV to above 500 meV. The computer controlled analyzer system offers six standard modes of operation, including a diffraction detector, a position-sensitive detector (PSD) in diffraction mode, horizontal energy focusing analyzer with detector, a Q-E mode employing a flat analyzer and PSD, a constant-E mode with the analyzer crystal system and PSD, and a conventional mode with a selection of Söller collimators and detector. Additional configurations for specific measurement needs are also available. This paper discusses the capabilities and performance for this new state-of-the-art neutron spectrometer.

Boron Vehiculating Nanosystems for Neutron Capture Therapy in Cancer Treatment.

Boron neutron capture therapy is a low-invasive cancer therapy based on the neutron fission process that occurs upon thermal neutron irradiation of 10B-containing compounds; this process causes the release of alpha particles that selectively damage cancer cells. Although several clinical studies involving mercaptoundecahydro-closo-dodecaborate and the boronophenylalanine-fructose complex are currently ongoing, the success of this promising anticancer therapy is hampered by the lack of appropriate drug delivery systems to selectively carry therapeutic concentrations of boron atoms to cancer tissues, allowing prolonged boron retention therein and avoiding the damage of healthy tissues. To achieve these goals, numerous research groups have explored the possibility to formulate nanoparticulate systems for boron delivery. In this review. we report the newest developments on boron vehiculating drug delivery systems based on nanoparticles, distinguished on the basis of the type of carrier used, with a specific focus on the formulation aspects.

Boron Neutron Capture Therapy: Current Status and Challenges.

Boron neutron capture therapy (BNCT) is a re-emerging therapy with the ability to selectively kill tumor cells. After the boron delivery agents enter the tumor tissue and enrich the tumor cells, the thermal neutrons trigger the fission of the boron atoms, leading to the release of boron atoms and then leading to the release of the α particles (4He) and recoil lithium particles (7Li), along with the production of large amounts of energy in the narrow region. With the advantages of targeted therapy and low toxicity, BNCT has become a unique method in the field of radiotherapy. Since the beginning of the last century, BNCT has been emerging worldwide and gradually developed into a technology for the treatment of glioblastoma multiforme, head and neck cancer, malignant melanoma, and other cancers. At present, how to develop and innovate more efficient boron delivery agents and establish a more accurate boron-dose measurement system have become the problem faced by the development of BNCT. We discuss the use of boron delivery agents over the past several decades and the corresponding clinical trials and preclinical outcomes. Furthermore, the discussion brings recommendations on the future of boron delivery agents and this therapy.

Boron Neutron Capture Therapy: Clinical Application and Research Progress.

Boron neutron capture therapy (BNCT) is a binary modality that is used to treat a variety of malignancies, using neutrons to irradiate boron-10 (10B) nuclei that have entered tumor cells to produce highly linear energy transfer (LET) alpha particles and recoil 7Li nuclei (10B [n, α] 7Li). Therefore, the most important part in BNCT is to selectively deliver a large number of 10B to tumor cells and only a small amount to normal tissue. So far, BNCT has been used in more than 2000 cases worldwide, and the efficacy of BNCT in the treatment of head and neck cancer, malignant meningioma, melanoma and hepatocellular carcinoma has been confirmed. We collected and collated clinical studies of second-generation boron delivery agents. The combination of different drugs, the mode of administration, and the combination of multiple treatments have an important impact on patient survival. We summarized the critical issues that must be addressed, with the hope that the next generation of boron delivery agents will overcome these challenges.

[Neutrons in Clinical Practice: BNCT].

Boron neutron capture therapy (BNCT) is a radiation therapy that uses charged particles produced by a nuclear reaction between thermal neutrons and 10B. A high-intensity neutron source is required to perform BNCT, and it is important to understand the behavior of neutrons. Since BNCT using accelerators has been approved as a medical device, the number of treatment facilities is expected to increase in the future. This article describes the basic knowledge required to understand BNCT in clinical practice, including neutron generation and material interactions, as well as radiation protection considerations specific to BNCT.

Intensity-modulated irradiation for superficial tumors by overlapping irradiation fields using intensity modulators in accelerator-based BNCT.

The distribution of the thermal neutron flux has a significant impact on the treatment efficacy. We developed an irradiation method of overlapping irradiation fields using intensity modulators for the treatment of superficial tumors with the aim of expanding the indications for accelerator-based boron neutron capture therapy (BNCT). The shape of the intensity modulator was determined and Monte Carlo simulations were carried out to determine the uniformity of the resulting thermal neutron flux distribution. The intensity modulators were then fabricated and irradiation tests were conducted, which resulted in the formation of a uniform thermal neutron flux distribution. Finally, an evaluation of the tumor dose distribution showed that when two irradiation fields overlapped, the minimum tumor dose was 27.4 Gy-eq, which was higher than the tumor control dose of 20 Gy-eq. Furthermore, it was found that the uniformity of the treatment was improved 47% as compared to the treatment that uses a single irradiation field. This clearly demonstrates the effectiveness of this technique and the possibility of expanding the indications to superficially located tumors.

Comparison of FANT results using the ENDF/B-VII.1, JEFF-3.3 and TENDL2017 nuclear data libraries.

FANT (Fuente Ampliada de Neutrones Térmicos; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, that has been developed by the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politecnica de Madrid (LMN-UPM). This device is a parallelepiped box made of high-density polyethylene (HDPE), moderator material, that uses an A95241m/B49e neutron source of 111 GBq nominal activity for irradiating materials. The facility design was previously optimized, and the neutron spectra were estimated by extensive calculations with the MCNP6.1 code and carrying out experimental measurements (Bedogni et al., 2017). The facility takes advantage of the scattering reactions of neutrons with the HDPE surfaces of the chamber, where the moderation process is effective, achieving relevant thermal neutron fluence rates. The main goal of this work has been to simulate and analyse the FANT system by Monte Carlo methods using the MCNP6.1 code, employing 3 different nuclear data libraries: ENDF/B-VII.1, JEFF-3.3 and TENDL 2017. The transport of thermal neutrons in HDPE, E < 1eV, has been calculated in all the cases taking into account the thermal S (α,ß) treatment. The results achieved in this work have been compared with those previously obtained in the former development of FANT, using the MCNP6.1 code with the ENDF/B-VII.1 nuclear data, and experimental measurements. These results have shown that the JEFF-3.3 nuclear data library is the nuclear data library that provides of the best matching between the MCNP computational results, and the experimental data collected at FANT. Hence, the JEFF-3.3 nuclear data library seems to be the most correct library to design and benchmark thermal neutron activation devices.

Experimental characterization of FANT, a new thermal neutron source.

FANT is the acronym of Enhanced Thermal Neutron Source (Fuente Ampliada de Neutrones Térmicos, in Spanish). This is a parallelepiped box of high-density polyethylene moderator and an isotopic neutron source. The moderator has a cylindrical irradiation chamber where a rather uniform thermal neutron flux is obtained. The FANT design was previously optimized and the neutron spectra were estimated by Monte Carlo calculations with the MCNP6.1 code. To check the characteristics of the FANT thermal neutron field, measurements have been performed at the reference point inside the irradiation chamber with a Bonner sphere spectrometer holding a small 6LiI(Eu) thermal neutron detector. To unfold the neutron spectrum BUNKIUT with UTA4 response matrix and NSDann Ver 4.0 codes were used. Some issues have been found and recommendations are made about the use of large BSS inside narrow spaces, and about the capacity of NSDann code to unfold these kind of spectra. However, the results confirm that the moderation process in FANT is very effective and allows obtaining useful thermal neutron fluence rates.