Fused deposition modeling (FDM) 3D printing is an additive manufacturing process capable of rapidly building three-dimensional computer-modeled objects. The technology offers an inexpensive and efficient technique to manufacture customized objects with intricate geometries using a simple printing process. However, FDM is currently restricted in application due to a limited availability of functional materials. Research in the field has focused on incorporating functional characteristics into printable polymers to expand application of FDM technology. In this work, neutron radiation shielding was targeted as an addition to FDM materials. By creating a composite material using a thermoplastic polymer matrix and boron nitride additive, neutron shielding of FDM-printed samples was enhanced from 50% attenuation in polymer specimens to 72% in composite specimens. The enhanced functionality of this new material enables FDM technology to be used in the manufacture of aerospace components, where neutron radiation presents a significant hazard.

Ceramic fiber–matrix composites (CFMCs) are exciting materials for engineering applications in extreme environments. By integrating ceramic fibers within a ceramic matrix, CFMCs allow an intrinsically brittle material to exhibit sufficient structural toughness for use in gas turbines and nuclear reactors. Chemical stability under high temperature and irradiation coupled with high specific strength make these materials unique and increasingly popular in extreme settings. This paper first offers a review of the importance and growing body of research on fiber–matrix interfaces as they relate to composite toughening mechanisms. Second, micropillar compression is explored experimentally as a high-fidelity method for extracting interface properties compared with traditional fiber push-out testing. Three significant interface properties that govern composite toughening were extracted. For a 50-nm-pyrolytic carbon interface, the following were observed: a fracture energy release rate of ∼2.5 J/m2, an internal friction coefficient of 0.25 ± 0.04, and a debond shear strength of 266 ± 24 MPa. This research supports micromechanical evaluations as a unique bridge between theoretical physics models for microcrack propagation and empirically driven finite element models for bulk CFMCs.

It is known that n-Si solar cells have higher efficiency than p-Si solar cells. One of the problems connected with n-Si application for solar cell production is the difficulty of using Czochralski method for growing n-Si ingots, uniform in structure. The present paper examines the possibility of production of n-Si ingots, uniform in resistance, by neutron transmutation doping (NTD) for photovoltaics using the mathematical modeling method. The provided calculation data are obtained by MCU-RFFI/A accounting code with DLC/MCUDAT-1.0 constant library developed by «Kurchatov Institute» Russian Research Center. The MCU accounting code is used for solution of the neutron-transport equation by Monte-Carlo procedure on the basis of estimated nuclear data for arbitrary three-dimensional geometry systems.

The present paper provides the estimation of uniformity of neutron-flux density along the ingot length and radius; dependence of silicon resistance on duration of irradiation. These studies established the neutron flux density distribution along the ingot length and radius; regularities of silicon resistance changes on duration and intensity of irradiation.

It is known that microstructure of metallic polycrystalline materials irradiated with neutrons is often characterized by a high degree of heterogeneity in distribution of radiation-induced defects. Depleted zones are located along grain boundaries and their width is not only determined by irradiation temperature and damage dose, but also by migration of point defects and dislocations integrity, that makes it more difficult to interpret experimental results of this phenomenon. At present, denuded zones are still objects for investigation as they influence both operation characteristics of reactor materials and their safe long-term storage. In this work, denuded zones in hexagonal ducts of spent fuel assemblies constructed from 0.08C-16Cr-11Ni-3Mo and 0.12C-18Cr-10Ni-Ti stainless steels from BN-350 fast nuclear reactor were investigated by TEM. There were determined some irradiation parameters affecting the development of denuded zones and their width; void size distributions in near-grain boundary regions are presented. There was shown redistribution of alloying elements at grain boundaries using Energy-dispersive X-ray spectroscopy (EDS).

One of the most important parts of a hybrid reactor is the cladding because it should withstand high temperatures, neutrons with high energy, high neutron flux, as well as provide the first security contention. Besides, the material should have good mechanical properties to remove the heat. Although, the cladding material choice will have a great influence on criticality calculations. In previous works on ADS there is no cladding used, therefore in this paper it is tested different cladding materials based on SS-316, ODS, T91 and 15-15Ti used in nuclear reactors, to study the variations on the fuel depletion and variations on the neutronic parameters. The results using the cladding are compared with the one obtained without using it. The best material choice is based on the neutronic parameters that presents the closest behavior to the ADS simulated without clad.

This paper studies the performance of (U, Pu)C fuel in a hexagonal assembly of a GFR (Gas Fast Reactor). The SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation version 6.0) code was used in the calculation. The goal is to evaluate the behavior of the infinite multiplication factor (kinf) for a heterogeneous assembly model using four nuclear data libraries: V6-238, V7-238, ENDF/B-VI.8 and ENDF/B-VII.0. The burnup of (U, Pu)C was performed by the TRITON-6 module, and the isotopic concentrations were evaluated during the cycle. The present work comprises calculations at Zero Power and Full Power condition. This study intends to achieve more information about different Fast Reactors.

This article reviews the application of nanomaterials for radiation shielding to protect humans from the hazards of radiation in space. The focus is on protection from space radiation, including galactic cosmic radiation (GCR), solar particle events (SPEs), and neutrons generated from the interactions of the GCR and SPEs with the intervening matter. Although the emphasis is on protecting humans, protection of electronics is also considered. There is a significant amount of work in the literature on materials for radiation shielding in terrestrial applications, such as for neutrons from nuclear reactors; however, the space environment poses additional and greater challenges because the incident particles can have high charges and extremely high energies. For materials to be considered for radiation shielding in space, they should perform more than just the radiation-shielding function; hence the emphasis is on multifunctional materials. In space, there is also the need for materials to be very lightweight and capable of surviving temperature extremes and withstanding mechanical loading. Nanomaterials could play a significant role as multifunctional radiation-shielding materials in space.

The Oxide Dispersion Strengthened (ODS) materials are potential candidates as cladding tubes for Sodium-cooled Fast Reactors. The nano-oxides are finely dispersed within the grains and confer excellent mechanical properties to these alloys. Hence, assessing nano-particle stability under irradiation remains crucial to guarantee safe use of these materials. Although neutron irradiation remains a binding and challenging experimental study to conduct, difficulties can be overcome by ion beam processing. Ion beam processing of the ODS material allows to identify the radiation-induced Ostwald ripening as the mechanism governing the nano-particle response under irradiation. The result is the increase in size and a decrease in density of the finely dispersed Y2Ti2O7 nano-particles. Under neutron irradiation, radiation-induced Ostwald ripening appears to be less effective since a slight growth of nano-particles is observed. Further, our approach shows that nanoparticle growth kinetics should scale as φ1/3, φ being the radiation flux. This suggests that the low irradiation flux is at the origin of the slower growth kinetics of the neutron irradiated particles. Both neutron and ion irradiation induce a modification of the nanoparticles/matrix interfaces which are generally flat and sharp prior to irradiation and present steps after irradiation. This could alter the nano-particle coarsening during irradiation.

Synchrotron-based x-ray techniques are used to bring complementary information to local probes such as atom probe tomography and transmission electron microscopy. Two examples of nuclear materials used for the cladding of fuel assembly are given: oxide dispersion strengthened (ODS) alloys and M5™ Zr-based alloys. In both cases, synchrotron radiation analyses bring original results concerning nanosized secondary phases: for M5™, radiation-enhanced precipitation of β-Nb precipitates has been evidence and the crystallographic structure (lattice parameter and Nb content) is reported for the first time and for irradiated ODS, the dissolution of larger oxides is evidenced while a finer distribution of complex Y–Ti–O oxides still acts as obstacles for dislocations.

The displacement per atom (dpa) has been a specific issue to evaluate the damage in the first wall of the Tokamak. Two different first wall materials were evaluated. In this study, MCNP5 code was used to obtain the neutron flux, the energy deposition and the main reaction rates, on the inboard and outboard first wall. The damage calculations were performed by the SPECTER code using the neutronic parameters obtained by MCNP5. The Tokamak reactor modeled has similar dimensions to the ITER. Tungsten and beryllium alloys were simulated on the outboard first wall. The results indicate which material has a higher resistance to be damage and dpa values for the analyzed material.

Integrity and lifetime of reactor pressure vessels are practically determined by their material resistance against fast/non-ductile failure and consequently by their radiation damage resulting in irradiation embrittlement and hardening. Mechanism of irradiation embrittlement of RPV materials depends on selected materials and operation conditions but their values depend in a large extent on reactor design, i.e. on neutron flux/fluence depending on RPV wall. Generally, new RPV design allows smaller neutron fluences but absolute value of irradiation embrittlement still depends on a choice of RPV material. Even though radiation damage (especially irradiation embrittlement) is important for RPV behavior, integrity and lifetime depends, in principle, on final value of applied fracture mechanics parameter –transition temperature. Thus, its initial value as well as its shift due to irradiation embrittlement is of interest but only the embrittlement can be affected during operation.

Novel and more conventional boron carbides were combined with n-type silicon to make heterojunction diodes, with neutron capture signal at zero applied bias. The boron carbides were based on the cross linking of closo-1,2-dicarbadodecaborane (ortho-carborane; 1,2-B10C2H12), and cross linking based on the combination of closo-1,2-dicarbadodecaborane (ortho-carborane; 1,2- B10C2H12) and pyridine. In the latter devices, pyridine concentration was varied; samples with a closo-1,2-dicarbadodecaborane (ortho-carborane; 1,2- B10C2H12) to pyridine ratio of 1:1 (BC:Py1) and 1:3 (BC:Py3). The result is a nonvolatile robust p-type semiconductor of boron carbide (B10C2Hx):(C5NHx)y. The I(V) curves for the resulting heterojunction diodes exhibit strong rectification where the normalized reverse bias leakage currents are largely unperturbed with increasing pyridine inclusion. The devices are largely gamma insensitive and yet neutron voltaic properties of these boron carbides is demonstrated. The neutron capture generated pulses from these heterojunction diodes were obtained at zero bias voltage although without the characteristic signatures of complete charge collection from boron neutron capture generated electron-hole pair production. These results, nonetheless, suggest that modifications to boron carbide may result in better neutron voltaic materials with linking groups chosen from family of aromatic compounds that stretch between borazine (B3N3H6) and benzene that point the way to a whole family of future studies that may ultimately lead to boron carbides better suited to low power and low flux neutron detection.

The purpose of this paper is to propose the enhancement of device detectors based on p-n junction in 4H-SiC for nuclear instrumentation. Particular emphasis is placed on the interest on using Boron isotope 10 as a Neutron Converter Layer in order to detect thermal neutrons. Here, we present the main results obtained during several irradiation tests performed in the Belgian Reactor 1 (BR1). We show the capability of our detectors by means of first results of the detector response at different reverse voltage biases and at different reactor power.

The effect of the micro-porous layer (MPL) in polymer electrolyte fuel cells (PEFCs) was studied by a combination of in situ visualization of the liquid water distribution and advanced electrochemical analysis using helox and O2 pulses. Four cells with and without MPLs on the anode and cathode side were tested. Visualization studies showed that the significant changes in performance observed when using an MPL on the cathode side cannot be related to a reduction of the water content in the cathode side diffusion layer (GDL). The helox/O2 pulse analysis indicated that two different mechanisms are responsible for the performance loss without an MPL.

In order to understand the vacancy behavior during incubation period before steady state void swelling, positron annihilation lifetime measurements was performed after isochronal annealing of austenitic stainless steel (Ti added modified SUS316SS) and ferritic stainless steel (F82H) irradiated by neutrons and electrons to a dose of 0.2 dpa. By electron and neutron irradiations below 363 K, vacancies and nano-voids containing of few vacancies were formed in both alloys. By increasing annealing temperatures, the lifetime decreased without forming nano-voids. The change of lifetime during the annealing indicated the formation and growth of staking fault tetrahedra (Ti added modified SUS316SS) and the annihilation of vacancies at precipitates (F82H).

Neutron detection scintillators based on rare-earth activated transparent glass and glass-ceramics are reported. Ce3+ doped gadolinium halides in 6LiF modified aluminosilicate glass matrices were synthesized by a melt-quench method followed by annealing. Their optical properties and alpha, neutron scintillation performance were investigated and compared to conventional 6Li-based scintillating glass.

Models of radiation growth proposed to date are all based on the assumption that the primary damage is produced by neutron irradiation in the form of single defects. These models do not account for the features of the cascade damage: intra-cascade clustering of self‑interstitial atoms (SIAs) and their one‑dimensional diffusion. During the last twenty years, a ‘Production Bias Model’ has been developed, which shows that the damage accumulation in cubic metals depends crucially on the cascade properties. The cascades in hcp zirconium are similar to those in cubic crystals; hence the model can provide a realistic framework for the hcp metals as well. In this work we present such a model in application to low-temperature (below 300°C) radiation growth.

Thin film gallium nitride (GaN) scintillators have been produced by MOCVD and made neutron-sensitive by applying an enriched lithium-6 fluoride (6LiF) conversion layer. The 6Li(n,α) reaction produces both alpha and triton particles, which have very penetration depths in GaN. The range and energy deposition characteristics of these particles in GaN have been simulated. Alpha-induced scintillation was measured in silicon-doped GaN using an americium-241 (241Am) source. The thermal neutron responses of the 6LiF-coated GaN scintillator were tested using two thermal neutron sources, an 241Am-Be source inside a graphite pile and a reactor source. The scintillator was found to have a linear response to thermal neutron flux level over a range of more than three orders of magnitude.

We propose a novel concept of optical detection of thermal neutrons in a passive device that exploits transmutation of Dy-164, a dominant, naturally occurring isotope of dysprosium, into a stable isotope of either holmium Ho-165 or erbium Er-166. Combination of the high thermal neutron capture cross section of ~2,650 barns and transmutation into two other lanthanides makes Dy-164 a very attractive alternative to traditional methods of neutron detection that will be completely insensitive to gamma irradiation, thus reducing greatly the likelihood of false alarms. The optically enabled neutron detection relies on significant differences in optical properties of Dy, Ho, and Er that are not sensitive to a particular isotope, but change considerably from one element to another. While the concept applies equally well to bulk materials and to nanocrystals, nanocrystalline approach is much more attractive due to its significantly lower cost, relative ease of colloidal synthesis of high quality nanocrystals (NCs), and superior optical and mechanical properties of NCs compared to their bulk counterparts. We report on colloidal synthesis of DyF3 NCs, both doped and undoped with Ho and co-doped with Ce and Eu to enhance their optical properties. We also report on DyF3:10%Ce and DyF3:10%Eu NCs irradiated with thermal neutrons from a Cf-252 source and their optical characterization.

Solid-state neutron detectors from heterostructures that incorporate Gd intrinsically or as a dopant may significantly benefit from the high thermal neutron capture cross section of gadolinium. Semiconducting devices with Gd atoms can act as a neutron capture medium and simultaneously detect the electronic signal that characterizes the interaction. Neutron capture in natural isotopic abundance gadolinium predominantly occurs via the formation of 158mGd, which decays to the ground state through the emission of high-energy gamma rays and an internal conversion electron. Detection of the internal conversion electron and/or the subsequent Auger electron emission provides a distinct and identifiable signature that neutron capture has occurred. Ensuring that the medium responds to these emissions is imperative to maximizing the efficiency and separating out other interactions from the radiation environment. A GEANT4 model, which includes incorporation of the nuclear structure of Gd, has been constructed to simulate the expected device behavior. This model allows the energy deposited from the decay of the meta-stable state to be localized and transported, providing for analysis of various device parameters. Device fabrication has been completed for Gd doped HfO2 on n-type silicon, Gd2O3 on p-type silicon and Gd2O3 on SiC for validation of the code. A preliminary evaluation of neutron detection capabilities of these devices using a GEANT4 modeling approach is presented.