Thermal Neutron Research Articles

Temperature-Dependent Elastic Properties of B4C from First-Principles Calculations and Phonon Modeling

Boron carbide plays a crucial role in various extreme environment applications, including thermal barrier coatings, aerospace applications, and neutron absorbers, because of its high thermal and chemical stability. In this study, the temperature-dependent elastic stiffness constants, thermal expansion coefficient, Helmholtz free energy, entropy, and heat capacity at a constant volume (Cv) of rhombohedral B4C have been predicted using a quasi-harmonic approach. A combination of volume-dependent first-principles calculations (density functional theory) and first-principles phonon calculations in the supercell framework has been performed. Good agreement between the elastic constants and structural parameters from static calculations is observed. The calculated thermodynamic properties from phonon calculations show trends that align with the literature. As the temperature rises, the predicted free energy follows a decreasing trend, while entropy and Cv follow increasing trends with temperature. Comparisons between the predicted room temperature thermal expansion coefficient (TEC) (7.54×10−6 K−1) and bulk modulus (228 GPa) from the quasi-harmonic approach and literature results from experiments and models are performed, revealing that the calculated TEC and bulk modulus fall within the established range from the limited set of data from the literature (TEC = 5.73–9.50 ×10−6 K−1, B = 221–246 GPa). Temperature-dependent Cijs are predicted, enabling stress analysis at elevated temperatures. Overall, the outcomes of this study can be used when performing mechanical and thermal stress analysis (e.g., space shielding applications) and optimizing the design of boron carbide materials for elevated temperature applications.

A microfluidic route for synthesis of scandium oxide microspheres, their characterization and neutron activation

Radioactive scandium-46 microspheres have applications in mapping flow in a chemical reactor through a technique known as radioactive particle tracking (RPT). In the present study a novel microfluidic method has been developed for synthesis of controlled size scandium oxide microspheres. An inline/in-situ mixing of the scandium precursor and gelling agents was implemented which makes the microfluidic platform amenable for truly continuous operation. Microspheres of size varying from 937 to 666 μm were produced by varying O/A ratio from 10 to 30. Perfectly spherical and monodispersed (PDI <10 %) microspheres were obtained at O/A 15 and beyond. The morphology, elemental composition, and structure of the microspheres were analysed by SEM, EDS and XRD, respectively. Subsequently the microspheres were irradiated with thermal neutrons in a nuclear reactor to obtain radioactive Sc-46 oxide microspheres. The activity produced on each Sc-46 microspheres with different sizes was in the range 19.5–34.0 MBq.

Activation level of the concrete building and pressure vessel in JAEA-Tokai tandem accelerator

ABSTRACT The activation level of the JAEA-Tokai tandem accelerator facility was investigated experimentally in advance of the future decommissioning. JAEA-Tokai tandem accelerator facility has a higher terminal voltage of 18 MV and a larger total floor area, Compared to other electrostatic accelerators with terminal voltage of 1 MV to 6 MV. Therefore, determination for ‘where,’ ‘what,’ and ‘how many’ nuclides are produced in the facility is crucial. Thermal neutrons generated by beam losses associated with accelerator operations contribute significantly to the activation of equipment and facilities. The accumulated activities of60Co and152Eu; the most considerable radionuclides at the decommissioning, can be deduced by the thermal neutron fluence rate during the accelerator operation. In this study, thermal neutron fluence measurement on the surface of the pressure vessel and concrete building was conducted with conventional methods using dosimeters and metal foil detectors as well as a new method using a portable γ-ray detector. The thermal neutron fluence in the facility during the accelerator operation ranges from 101 to 104 n/cm2/s. The sum of deduced activities of60Co and152Eu in 50 years is much lower than the clearance level of 0.1 Bq/g in all areas except in the irradiation room.

Response of LaBr2.85Cl0.15:Ce, LaBr[formula omitted]:Ce and NaI:Tl crystals to fast and thermal neutrons

This communication reports the measurement of direct response for fast and thermal neutrons of a newly marketed scintillation crystal LaBr2.85Cl0.15:Ce, along with the widely used LaBr3:Ce and NaI:Tl crystals. It continues our investigation on the possible utility of LaBr2.85Cl0.15:Ce crystal for the detection of high-energy γ-rays and neutrons. This is the first report of the neutron response of the LaBr2.85Cl0.15:Ce crystal. The neutron response was measured for the three detectors using a calibrated Am-Be source, which emits both neutrons and γ-rays. To eliminate the events due to the interaction of the 4.4 MeV γ-ray emitted from Am-Be, the γ-ray response of all three crystals was simulated using the Monte Carlo simulation toolkit GEANT4. Events due to 2.22 MeV γ-ray produced by the capture of thermal neutrons in the paraffin moderator were also eliminated with the aid of simulations. In addition, the response due to fast neutrons was also simulated using GEANT4 and compared with the measured spectrum. This is the first report of the simulation of fast neutron response of LaBr2.85Cl0.15:Ce using GEANT4. The fast neutron detection efficiency of the three detectors has been obtained from the measured spectra.

Preparation of [64Cu]Cu–NOTA complex as a potential renal PET imaging agent using 64Cu produced via the direct activation route

Abstract Glomerular filtration rate (GFR) could be determined more accurately using renal positron emission tomography (PET) than conventional gamma imaging. Copper-64 [T ½ = 12.7 h, E β+ (max) = 653 keV, β+ branching ratio = 17.8 %, 1346 keV γ-photon (0.54 %), EC (43.8 %), β− emission (38.4 %)] in the form of its hydrophilic complex with 1,4,7-triazacyclononane-1,4,7-triacetic acid (NOTA) is proposed as a potent formulation for renal PET imaging. A lyophilized kit was developed for formulation of ∼370 MBq dose of [64Cu]Cu–NOTA complex in a facile single step process using 64Cu produced by thermal neutron activation in a research reactor. The complex could be synthesized with &gt;99 % yield and retained its integrity even when challenged by apoferritin. The rapid accumulation of [64Cu]Cu–NOTA in the kidney and clearance through urinary path was demonstrated using PET/CT imaging and ex vivo biodistribution study carried out in healthy Wistar rats to elucidate its effectiveness as a renal PET-imaging agent.

Application of thin boron deposit by electrophoresis as neutron detectors

Abstract Detecting nuclear radiation presents a distinctive challenge, particularly with neutrons, which are neutral particles. The method of direct detection involves the utilization of a converter material, acting as an intermediary. Boron plays a pivotal role in this process, reacting with thermal neutrons to generate alpha particles and lithium, with a notable energy release of 2.314 MeV during the 10B (n,α) 7Li reaction. This facilitates effective identification and measurement of neutrons in radiation detection systems. The paths of the particles α (for E = 1.474 MeV) and Li (for E Li = 0.842 MeV). The active medium of the nuclear detector, typically a gas, undergoes ionization by these highly charged particles, or they form ion pairs that are subsequently collected by electrodes to produce the signal at the detector’s output. Various deposit methods can be used for this purpose, electrophoresis offers a distinct advantage in terms of both simplicity and precision. This study details the utilization of the electrophoresis technique for the deposition of boron on the tube walls of prototype detectors developed within our laboratory.

Design of moderator and collimator based on compact D-T neutron source for neutron imaging

ABSTRACT The Compact thermal neutron imaging systems have become an important development trend. Compact thermal neutron imaging systems can use neutron tubes as neutron sources because of their low cost, small size, and portability. In order to achieve optimal imaging quality, the moderator and collimator in a neutron imaging system based on neutron tubes source are designed and simulated using SuperMC. The entrance position of the collimator is determined by analyzing the performance of different materials with their parameter’s optimization. It is suggested that a cylinder-shaped structure made of carbon, hydrogenous materials, and heavy metals can significantly increase the thermal neutron flux at the sample location. Further dimension optimization is carried out to improve the neutron beam collimation performance. A vertically divergent collimator with a divergence angle of 5° and an air gap is selected, along with the addition of a B4C lining material. According to the modelling results, the sample position can be as far away from the neutron source as 77 cm with a neutron source yield of 1.14 × 109 n/s. The collimation ratio is 29, the thermal neutron flux is 1.08 × 103 n/cm2·s, and n/γ is 1.44 × 1012 n/cm2·Sv, which is suitable for thermal neutron radiography.

Gadolinium Concentration Measurement with an Atomic Absorption Spectrophotometer

Abstract Because gadolinium (Gd) has the highest thermal neutron capture cross section, resulting in an 8 MeV gamma cascade upon capture, it has been proposed for dissolution in water Cherenkov detectors to achieve efficient neutron tagging capabilities. Whereas metallic Gd is insoluble in water, several compounds are very easy to dissolve. Gadolinium sulfate, Gd2(SO4)3, has been thoroughly tested and proposed as the best candidate. Accurate measurement of its concentration, free of doubt from impurities in water, is crucial. An atomic absorption spectrophotometer (AAS) is a device that suits this purpose and is widely used to measure the concentration of many elements. In this study, we describe three different approaches to measure Gd sulfate concentrations in water using an AAS: doping samples with potassium and lanthanum, and employing tantalum and tungsten platforms.

MCNP simulation and experimental study in situ low-grade copper analysis based on PGNAA

A prompt gamma neutron activation analysis (PGNAA) facility utilizing 252Cf source has been developed for in situ analysis of copper samples. Monte Carlo simulation is employed to determine optimal sizes for neutron moderator, gamma-ray shielding material, and thermal neutron absorber. Subsequently, based on the parameters optimized by MCNP, the PGNAA facility was constructed. Five sets of experimental samples containing low-grade copper concentration of 0 %, 0.5 %, 1 %, 1.5 % and 2 % are measured with the PGNAA facility. The results show that the minimum detectable concentration of copper is 0.218 %. The maximum relative deviation of copper is 8.53 %.

Development of a Triple-GEM detector with strip readout and GEMINI chip for X rays and neutron imaging

Thermal neutron imaging can be a useful tool in the study of the internal structure of an object. The different attenuation properties of the materials with respect to X rays give rise to different interactions and the result is a complementary non-destructive analysis, which can provide important additional information. This technique has been successfully employed in different areas of work, especially in material science and cultural heritage studies. This paper describes the development of a new detection system and its characterization performed with X ray emissions. The system features the use of a gaseous detector, based on the Gas Electron Multiplier technology, and a fully digital electronic readout, with a combination of custom-made ASICs (called GEMINI) and FPGA boards, enabling fast single photon counting. The detector can be thus used directly for X ray imaging, while the addition of a suitable converter in its active volume will allow for detection of neutrons and for reconstruction of their tracks. The readout system is based on a x-y strip structure and features the reconstruction of single events through the center of mass methodology, allowing for accurate tomography, with sub-mm spatial resolution, in combination with sub-ms time resolution and high rate capabilities (up to MHz/mm2).

A two-layer Timepix3 stack for improved charged particle tracking and radiation field decomposition

Abstract We characterize a novel instrument designed for radiation field decomposition and particle trajectory reconstruction for application in harsh radiation environments. The device consists of two Timepix3 assemblies with 500 µm thick silicon sensors in a face-to-face geometry. These detectors are interleaved with a set of neutron converters: 6LiF for thermal neutrons, polyethylene (PE) for fast neutrons above 1 MeV, and PE with an additional aluminum recoil proton filter for neutrons above ∼4 MeV. Application of the coincidence and anticoincidence technique together with pattern recognition allows improved separation of charged and neutral particles, their discrimination against γ-rays and assessment of the overall directionality of the fast neutron field. The instrument's charged particle tracking and separation capabilities were studied at the Danish Center for Particle Therapy (DCPT), the Proton Synchrotron, and Super Proton Synchrotron with protons (50–240 MeV), pions (1–10 GeV/c and 180 GeV/c). After developing temporal and spatial coincidence assignment methodology, we determine the relative amount of coincident detections as a function of the impact angle, present the device's impact angle resolving power (both in coincidence and anticoicidence channels). The detector response to neutrons was studied at the Czech Metrology Institute (CMI), at n_ToF and the Los Alamos Neutron Science Center (LANSCE), covering the entire spectrum from thermal up to 600 MeV. The measured tracks were assigned to their corresponding neutron energy by application of the time of flight technique. We present the achieved neutron detection efficiency as a function of neutron kinetic energy and demonstrate how the ratio of events found below the different converters can be used to assess the hardness of the neutron spectrum. As an application, we determine the neutron content within a PMMA phantom just behind the Bragg-peak during clinical irradiation condition with protons of 160 MeV.

Fabrication and Properties for Thermal Neutron Detection of 6LiCl/Rb2CeCl5 Eutectic Scintillator

The 3He gas is commonly used for the detection of thermal neutrons. However, with the depletion of 3He gas, there is a need to develop new solid scintillators for thermal neutron detection. Solid scintillators containing 6Li, which have large neutron capture cross-sections and a large amount of energy released by transmutation reactions, are commonly used as alternative candidates. However, only single-crystal scintillators are currently used, and their 6Li concentration is limited by their chemical composition. In this study, we designed, grew, and evaluated a new eutectic scintillator, Rb2CeCl5/LiCl, which can improve the 6Li concentration compared with single-crystal scintillators. Rb2CeCl5, which was selected as the scintillator phase, has excellent scintillator properties (light yield: 36,000 photons/MeV, decay time: mostly 24 ns, slightly 153 ns), and is less deliquescent than other halide scintillators. The crystal grown using the vertical Bridgman method exhibited a eutectic phase composed of Rb2CeCl5 and LiCl. The eutectic crystals exhibited Ce3+ 5d-4f emissions, with a peak between 360 and 370 nm. The Rb2CeCl5 phase was identified as the luminescent phase via cathodoluminescence mapping, and 16,000 photons/neutron of the light yield and 56.1 ns of the decay time were observed. This study indicates that the Rb2CeCl5/LiCl eutectic scintillator is a promising candidate for use in thermal neutron detectors.

Boron‐polymer composites engineered for compression molding, foaming, and additive manufacturing

AbstractBoron (specifically 10B) is the element of choice to shield thermal neutrons due to its large (n, α) cross‐section; however, very few polymer composites containing high boron concentrations are available. This study aimed to determine the maximum possible amount of boron that could be introduced into a polymer matrix. Diverse manufacturing techniques, ranging from additive manufacturing to compression molding, were employed to fabricate inks and filaments for 3D printing, foams, and flexible pads. Composites using siloxanes, poly(lactic acid), and acrylonitrile butadiene styrene containing up to 80 wt% boron were sucessufully fabricated. The addition of known plasticizers (polyethylene glycol) and reinforcing agents (carbon nanofibers and fumed silica) helped to overcome fabrication problems such as clogging of the printing nozzle or crumbling of compression molded parts. In addition, the thermal‐mechanical properties of these novel boron composites were determined and shown to vary according to boron concentration, presence of additives, and fabrication techniques utilized.

INTERVAL MODEL FOR ASSESSING THE PURITY OF NUCLEAR STRUCTURAL METALS BY MICROHARDNESS

The article considers the possibilities of evaluating the purity of Zr-1%Nb zirconium alloy samples – the main structural material of the thermal neutron reactors active zones with a water coolant – according to the microhardness of the metal. The correctness of the application of interval regression analysis is shown, if there is uncertainty, incompleteness of information, measurement errors in the experimental data for determining the oxygen content in metal samples and its microhardness. The method of estimating parameters of interval linear regression is presented.

Comparative Radiation Response of GaN and Ga2O3 Exposed to Ground-Level Neutrons

In this work, the radiation response of bulk GaN and Ga2O3 materials exposed to ground-level neutrons is studied by Geant4 numerical simulation, considering the whole atmospheric neutron spectrum at sea level, from thermal to high energies (GeV). The response of the two materials is compared in terms of the number and type of interactions and the nature of the secondary products produced, particularly in nuclear reactions. Our results highlight the importance of 14N(n,p)14C neutron capture in the radiation response of GaN, leading to large differences in the behavior of the two materials in terms of susceptibility to thermal and intermediate-energy (below 1 MeV) neutrons.

90Y]Yttria Alumino Silicate Glass Microspheres: A Biosimilar Formulation to "TheraSphere" for Cost-Effective Treatment of Liver Cancer.

Background: Selective internal radiation therapy (SIRT) using a suitable β--emitting radionuclide is a promising treatment modality for unresectable liver carcinoma. Yttrium-90 (90Y) [T1/2 = 64.2 h, Eβ(max) = 2.28 MeV, no detectable γ-photon] is the most preferred radioisotope for SIRT owing to its favorable decay characteristics. Objective: The present study describes indigenous development and evaluation of intrinsically radiolabeled [90Y]yttria alumino silicate ([90Y]YAS) glass microsphere, a formulation biosimilar to "TheraSphere" (commercially available, U.S. FDA-approved formulation), for SIRT of unresectable liver carcinoma in human patients. Methods: YAS glass microspheres of composition 40Y2O3-20Al2O3-40SiO2 (w/w) and diameter ranging between 20 and 36 μm were synthesized with almost 100% conversion efficiency and >99% sphericity. Intrinsically labeled [90Y]YAS glass microspheres were produced by thermal neutron irradiation of cold YAS glass microspheres in a research reactor. Subsequent to in vitro evaluations and in vivo studies in healthy Wistar rats, customized doses of [90Y]YAS glass microspheres were administered in human patients. Results: [90Y]YAS glass microspheres were produced with 137.7 ± 8.6 MBq/mg YAS glass (∼6800 Bq per microsphere) specific activity and 99.94% ± 0.02% radionuclidic purity at the end of irradiation. The formulation exhibited excellent in vitro stability in human serum and showed >97% retention in the liver up to 7 d post-administration when biodistribution studies were carried out in healthy Wistar rats. Yttrium-90 positron emission tomography scans recorded at different time points post-administration of customized dose of [90Y]YAS glass microspheres in human patients showed near-quantitative retention of the formulation in the injected lobe. Conclusions: The study confirmed the suitability of indigenously prepared [90Y]YAS glass microspheres for clinical use in the treatment of unresectable hepatocellular carcinoma.

Comparative analysis on small modular reactor (SMR) with uranium and thorium fuel cycle

The utilization of Small Modular Reactors (SMRs) and micro-reactors holds significant potential advantages. The primary benefits include their smaller size, accessibility to remote regions, and reduced land requirements. One noteworthy SMR is the NuScale Power Module (NPM), which has a thermal capacity of 160 MWt and operates on a 2-year cycle. However, there is a growing interest in exploring alternative fuels to enhance SMR performance and safety, and thorium has emerged as an appealing option due to its superior neutron characteristics and potential for higher fuel efficiency. This research aims to do preliminary research into adapting fuel for a thorium fuel cycle in the integral PWR (iPWR). The thorium cycle excels with an initial neutron multiplication factor of 1.134, almost the same as uranium (1.129). Additionally, the thorium cycle boasts superior neutron flux and Conversion Ratio (CR) in the thermal neutron energy range, enhancing sustainability with an average CR of 0.765, compared to 0.574 and 0.692 for uranium and uranium–thorium cycle. The addition of thorium improves efficiency in converting fertile material into fissile material with an enough criticality, where fuels incorporating thorium have a higher CR (Conversion Ratio). Furthermore, thorium fuel cycle has significantly lower production of plutonium isotopes compared to other fuel cycles. However, it faces a challenge with a high power peaking factor (PPF) of 1.923 at the Beginning of Cycle (BOC), requiring further research for mitigation.

Mathematical formulae for neutron self-shielding properties of media in an isotropic neutron field

In the current study, an ab initio derivation of the neutron self-shielding factor to solve the complex neutron transport problem of the decrease of the neutron flux as it penetrates into a material placed in an isotropic neutron field having equal flux in all directions. The theory of steady-state neutron transport was employed, starting from Stuart’s formula, to derive simple analytical formulae based on the integral cross-section parameters. The formulae could be adopted by the user according to various variables, such as the neutron flux distribution and geometry of the simulation at hand. The concluded formulae of the self-shielding factors comprise an inverted sigmoid function normalized with a weight representing the ratio between the macroscopic total and scattering cross-sections of the medium. The general convex volume geometries are reduced to a set of chord lengths, while the neutron interaction probabilities within the volume are parameterized to the epithermal and thermal neutron energies. The arguments of the inverted-sigmoid function were derived from a simplified version of neutron transport formulation. The derived analytic formulae agreed greatly with the experimental observations for different elements and geometries.

Evaluation of Gamma-Ray Dose Distribution in the Toroidal Direction of LHD Vacuum Vessel from Radionuclides Generated in Deuterium Plasma Experiments

The evaluation of activities from radionuclides inside a vacuum vessel is required for the safe maintenance and estimation of the activation level of fusion devices. The Large Helical Device (LHD), Toki, Japan, has been performing deuterium plasma experiments since 2017, and radionuclides have accumulated in components such as vacuum vessels. In this work, the gamma-ray spectrum and activity of each detected radionuclide in an LHD vacuum vessel were evaluated using portable high-purity germanium (P-HPGe) and numerical calculations of photon transport for comparison with the removed armor tiles to evaluate the occupational exposure in the maintenance of LHD experiments. Measurements using the P-HPGe detector were performed at 20 positions along the toroidal direction of the LHD, at each O port, and between each pair of adjacent O ports. Using the estimated efficiency of the photoelectronic effects inside the P-HPGe detector calculated by Monte Carlo N-particle code version 6.1, the activities of the radionuclides in the armor tiles attached to the vacuum vessel were evaluated. The estimated activity concentrations were compared with the measured activities of the armor tiles removed from the vacuum vessel. Gamma rays from 58Co, 54Mn, and 60Co were detected, and the measured activities of 58Co and 54Mn generated by fast neutrons in the armor tile were almost consistent with those obtained from the removed armor tiles. The radioactivity of 60Co in the armor tile was underestimated compared with the activity obtained from the removed armor tiles. The measurement of activation by fast neutrons is feasible, whereas activation by thermal neutrons requires more accurate calculations and surveys to be measured using this method.

Neutron flux evaluation algorithm with a combination of Monte Carlo and removal-diffusion calculation methods for boron neutron capture therapy.

In Japan, the clinical treatment of boron neutron capture therapy (BNCT) has been applied to unresectable, locally advanced, and recurrent head and neck carcinomas using an accelerator-based neutron source since June of 2020. Considering the increase in the number of patients receiving BNCT, efficiency of the treatment planning procedure is becoming increasingly important. Therefore, novel and rapid dose calculation algorithms must be developed. We developed a novel algorithm for calculating neutron flux, which comprises of a combination of a Monte Carlo (MC) method and a method based on the removal-diffusion (RD) theory (RD calculation method) for the purpose of dose calculation of BNCT. We present the details of our novel algorithm and the verification results of the calculation accuracy based on the MC calculation result. In this study, the "MC-RD" calculation method was developed, wherein the RD calculation method was used to calculate the thermalization process of neutrons and the MC method was used to calculate the moderation process. The RD parameters were determined by MC calculations in advance. The MC-RD calculation accuracy was verified by comparing the results of the MC-RD and MC calculations with respect to the neutron flux distributions in each of the cubic and head phantoms filled with water. Comparing the MC-RD calculation results with those of MC calculations, it was found that the MC-RD calculation accurately reproduced the thermal neutron flux distribution inside the phantom, with the exception of the region near the surface of the phantom. The MC-RD calculation method is useful for the evaluation of the neutron flux distribution for the purpose of BNCT dose calculation, except for the region near the surface.