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Children spend most of their time at school and participate in many activities that have the potential for causing injury. This study aims to describe the nationwide epidemiology of pediatric trauma sustained in school settings in the United States. In the 3-y analysis of 2017-2019 American College of Surgeons-Trauma Quality Program, all pediatric trauma patients (≤18 y) injured in a school setting were included and stratified based on place of injury, into elementary, middle, and high school (HS) groups. Descriptive statistics and multivariable logistic regression analysis were performed to identify the independent predictors of intentional injuries. 23,215 pediatric patients were identified, of which 15,264 patients were injured at elementary (57.6%), middle (17.5%), and high (25%) schools. The mean age was 9.5 y, 66.9% were male, 63.9% were white, the median injury severity score was 2 [1-4], and 95.6% had a blunt injury. Elementary school students were more likely to sustain falls (85%) and humerus fractures (43%) whereas HS students were more likely to be injured by assaults (17%). Overall, 7% of the students sustained intentional injuries. On multivariable logistic regression, male gender (odds ratio [OR] 1.54), Black race (OR 2.94), American Indian race (OR 1.88), Hispanic ethnicity (OR 1.77), positive drug screen (OR 4.9), middle (OR 5.2), and HSs (OR 10.6) were identified as independent predictors of intentional injury (all P < 0.01). Injury patterns vary across elementary, middle, and HSs. Racial factors appear to influence intentional injuries along with substance abuse. Further studies to understand these risk factors and efforts to reduce school injuries are warranted to provide a safe learning environment for children. [ABSTRACT FROM AUTHOR]

This work presents a discussion on a series of finite element analyses that assess stress evolution in the coating layers of tristructural isotropic (TRISO) particles in contact with each other while embedded in a matrix. The initial simulations were of applied uniaxial pressure versus matrix elastic modulus. These simulations predicted increasing stress in the silicon carbide coating layers of the TRISO particles with decreasing matrix elastic modulus. The second set of simulations focused on the effects of heating and cooling and the associated dimensional change on the state of stress in the coating layers. The general finding was that there was no significant difference below the coating layer's deposition temperature. However, above the deposition temperature, the contacting particles had higher stress compared with those that were separated. The third set of simulations focused on the effects of irradiation, specifically, creep, dimensional change, and swelling. An interface debonding model was introduced since these potential effects have a significant bearing on predicted stresses. [ABSTRACT FROM AUTHOR]

The intrinsic compatibility of silicon carbide (SiC) and hydrogen (H2) at high temperatures (2000‐2473 K) and pressure near one atmosphere was evaluated through a combination of thermodynamic calculations and hot hydrogen exposure testing. Thermodynamic calculations predict the decomposition of SiC in a hydrogen environment to form free silicon (Si) and free carbon (C). Free Si is predicted to vaporize from the surface as a volatile species, while free C may interact with H2 to form the hydrocarbons CH4 (T < 2100 K) or C2H2 (T > 2100 K). Coupons of high purity chemical vapor deposition (CVD) β‐SiC were exposed to slowly flowing hydrogen at temperatures ranging between 2000 and 2473 K. SiC experienced active attack as the result of H2 exposure, exhibiting linear weight loss kinetics and an Arrhenius dependence of weight loss on exposure temperature. The linear volatilization constant was experimentally evaluated to correspond with an activation energy of 370 ± 18 kJ/mol. Due to the dependence of observed corrosion rates on gas velocity, corrosion of SiC in flowing H2 was determined to be governed by external mass transfer of volatile Si species through the boundary layer. Experimentally derived mass losses were in good agreement with mass losses predicted by a boundary layer limited gas diffusion model. [ABSTRACT FROM AUTHOR]

Background: Humans are exposed to a variety of metals on a daily basis, and nickel is the most frequent contact allergen. Little is known about the frequency of sensitization to indium and iridium. Objectives: Study the prevalence of indium and iridium sensitization and evaluate the optimal patch test conditions. Methods: A total of 364 patients were patch tested at the allergy unit of the University Hospital of Basel. Pure metals, metal chlorides, and metal sulfates were applied in petrolatum or water in Inert Quadrate (IQ) test chambers for 2 days and read twice at day (D) 2, and between D4 and D7. Results: Eleven patients reacted to indium salts (3.0%), 13 to iridium salts (3.6%), and one reacted to both salts. None of the patients reacted to pure metals. Nineteen of the 23 patients who reacted either to indium or iridium showed concomitant positive reactions to other metals, mainly nickel and palladium. Conclusion: This retrospective clinical study provides insight into the prevalence and test conditions of two rarely tested metal allergens in a large patient cohort. A considerable number of indium‐ or iridium‐positive subjects had co‐sensitization to other metals. [ABSTRACT FROM AUTHOR]

A method for advanced manufacturing of silicon carbide offering complete freedom in geometric complexity in the three‐dimensional space is described. The method combines binder jet printing and chemical vapor infiltration in a process capable of yielding a high‐purity, fully crystalline ceramic—attributes essential for ideal performance in very high‐temperature applications or in the presence of displacement damage. Thermal conductivity and characteristic equibiaxial flexural strength of the resulting monolithic SiC at room temperature are 37 W·(m·K)−1 and 297 MPa, respectively. [ABSTRACT FROM AUTHOR]

The effects of steam pressure, velocity, and composition on SiC oxidation kinetics were studied. Pressure effects were tested at 1200°C from 0.1 to 1.4 MPa at a steam velocity of 0.25 cm/s. Velocity effects were tested in two furnaces at 0.45 MPa, 1200°C and 0.1 MPa, 1600°C with velocities ranging from 0.25 to 137 cm/s. Steam composition was altered by changing the reaction vessel material. Oxide morphology and composition were determined using optical and electron microscopy, and X‐ray diffraction. Porous oxides were observed whenever structural SiC from the reaction vessel saturated the steam with volatilized silica, H2, and CO. Oxidation kinetics were calculated by the change in SiC thickness. The steam velocity/recession rate followed a power‐law relationship of ~ 0.35 while the steam pressure/recession rate followed a power‐law relationship of ~ 1.78. [ABSTRACT FROM AUTHOR]

Irradiation-induced creep is one of the key material properties considered in designing structural components for nuclear reactors. This paper presents results for in situ irradiation-induced creep of chemical vapor–deposited 3C silicon carbide studied by instrumented irradiation in the Halden reactor in Norway. The specimens examined were irradiated at 300 °C and up to 2.5 × 1024 n/m2 (E > 0.1 MeV) under uniaxial tensile stress of <5 or 100 MPa. Irradiation-induced creep strain was defined as the differential time-dependent strain between the two specimens. Based on the dimensional inspections before and after irradiation, an axial primary creep strain of 0.06% was obtained at the end of irradiation. The lattice constant precisely determined from high-energy x-ray diffraction analysis showed a lattice expansion roughly accounting for the primary irradiation creep strain. Analysis of data from this and previous studies indicates that creep strain is significantly dependent on at least one of the experimental conditions, such as loading mode, neutron spectrum/flux, and material grade. [ABSTRACT FROM AUTHOR]

Abstract Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box. [ABSTRACT FROM AUTHOR]

The mechanical properties of a well-characterized irradiated light water reactor fuel pin with an average burnup of 72 MWd/kgU were investigated by utilizing nano-indentation technique. Young's modulus and hardness are reported as functions of radial positions covering from mid radial position in the urania pellet to high burnup structure, fuel-cladding chemical interaction zone and zirconium-based cladding. By probing microstructure of the high burnup zone, the mechanical properties of sub-grains and the interaction between them were examined at the nanometer scale. Force-displacement curves from each of these individual sub-grains suggest they resemble single crystal UO 2 . However, these restructured grains are weakly bonded to their neighbors. [ABSTRACT FROM AUTHOR]

Fully ceramic microencapsulated (FCM) fuel is a proposed fuel type for improved accident performance in LWRs (Light Water Reactors) that involves TRISO (TRistructural-ISOtropic) particles embedded in a nano-powder sintered silicon carbide (SiC) matrix. The TRISO particles contain a spherical fuel kernel ranging from 500 to 800 µm in diameter. The kernel and buffer layer are then coated with three layers, each of which is 30–40 µm thick, composed of dense inner pyrolytic carbon (IPyC), chemically vapor deposited silicon carbide (SiC) layer, and an outer pyrolytic carbon (OPyC) layer. These TRISO particles are then embedded in a fully dense sintered SiC matrix with an expected particle packing fraction of about 35–40% by volume. As is the case for gas reactor applications, the release of radioactivity into the coolant is dependent on the integrity of the silicon carbide layer of the TRISO particles, in addition to the SiC matrix. In this work, we report on fuel performance modeling of TRISO-bearing FCM fuel using the BISON code to simulate the thermo-mechanical behavior of this fuel in a prototypic LWR environment. This paper considers the effects of embedding a TRISO particle in the SiC pellet matrix and includes a discussion of the irradiation-induced dimensional change in the pyrolytic carbon (PyC) layers of the TRISO particle. Additionally, methods were developed to simulate a FCM pellet containing a large number of discrete and independent particles. Future work will report on developing an interface debonding model, a fracture model, and a radionuclide transport model. [ABSTRACT FROM AUTHOR]

The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber–reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry. [ABSTRACT FROM AUTHOR]

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd x Si (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions. [ABSTRACT FROM AUTHOR]

SiC/SiC composites are considered among leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. Under a large heat flux application (i.e. fuel cladding), the non-uniform changes in the dimensions and physical properties will lead to build-up of stresses in the structure over the course of time. To ensure reliable and safe operation of such a structure it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and elevated temperature. In this work, the foundation for 3D thermo-mechanical analysis of SiC/SiC cladding is put in place and a set of analyses with simplified boundary conditions has been performed. The analyses were carried out with two different codes that were benchmarked against one another and prior results in the literature. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These robust models may now be modified to take into account the axial and circumferential variation in neutron and heat flux to fully account for 3D effects. The results from the simple analyses show the development of high tensile stresses especially in the circumferential and axial directions at the inner region of the cladding. Based on the results obtained, design guidelines are recommended. For lack of certainty in or tailor-ability for the physical and mechanical properties of SiC/SiC composite material a sensitivity analysis is conducted. The analysis results establish a precedence order of the properties based on the extent to which these properties influence the temperature and the stresses. [ABSTRACT FROM AUTHOR]

Five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700 °C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Thermal resistivity increase as a result of irradiation was ∼20% higher when compared to CVD-SiC. [ABSTRACT FROM AUTHOR]

Background: Healthcare coverage has been shown to have implications in the prevalence of coronary artery disease. We explore the impact of lack of healthcare coverage on ischemic heart disease (IHD) mortality in the US. Methods: We obtained county-level IHD mortality and healthcare coverage data from the CDC databases for a total of 3,119 US counties. The age-adjusted prevalence of current lack of health insurance among individuals aged 18 to 64 years were obtained for the years 2018 and 2019 and were placed into four quartiles. First (Q1) and fourth quartile (Q4) had the least and highest age-adjusted prevalence of adults without health insurance, respectively. IHD mortality rates, adjusted for age through the direct method, were obtained for the same years and compared among quartiles. Ordinary least squares (OLS) regression for each demographic variable was conducted with the quartiles as an ordinal predictor variable and the age-adjusted mortality rate as the outcome variable. Results: We identified a total of 172,942 deaths related to ischemic heart disease between 2018 and 2019. Overall AAMR was higher in Q4 (92.79 [95% CI, 92.35–93.23]) compared to Q1 (83.14 [95% CI, 82.74–83.54]), accounting for 9.65 excess deaths per 100,000 person-years (slope = 3.47, p = 0.09). Mortality rates in Q4 for males (126.20 [95% CI, 125.42–126.98] and females (65.57 [95% CI, 65.08–66.05]) were higher compared to Q1 (115.72 [95% CI, 114.99–116.44] and 57.48 [95% CI, 57.04–57.91], respectively), accounting for 10.48 and 8.09 excess deaths per 100,000 person-years for males and females, respectively. Similar trends were seen among Hispanic and non-Hispanic populations. Northeastern, Southern, and Western regions had higher AAMR within Q4 compared to Q1, with higher prevalence of current lack of health insurance accounting for 49.2, 8.15, and 29.04 excess deaths per 100,000 person-years, respectively. Conclusion: A higher prevalence of adults without healthcare coverage may be associated with increased IHD mortality rates. Our results serve as a hypothesis-generating platform for future research in this area. [ABSTRACT FROM AUTHOR]

Metal Matrix Micro-encapsulated (M3) fuel consists of TRISO coated fuel particles directly dispersed in a matrix of zirconium metal to form a solid rod. In this integral fuel concept the cladding tube and the failure mechanisms associated with it have been eliminated; therefore, M3 fuel, compared to existing fuel designs, is expected to provide greatly improved operational performance. The main challenge to the deployment of M3 fuels is the low heavy metal load. This needs to be compensated with high density fuel (uranium nitride), larger rods and mainly higher enrichment. This study found that M3 fuel requires about 15.5% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in a PWR. In order to achieve comparable reactivity feedbacks, in particular moderator temperature, the pitch to rod diameter ratio should be reduced to 1.20 compared to the typical 1.326. As M3 fuel provides a better path for heat transfer from the fuel, such smaller pitch is expected to be feasible, but further evaluations especially to examine thermal-hydraulic aspects are required. Analyses of possible load patterns for burnable poisons showed that M3 fuel maintains similar cycle length to that of standard fuel and pin power peaks are also comparable. Nevertheless, further optimization will be required to limit assembly power peaks. [ABSTRACT FROM AUTHOR]

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH 1.6 ) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup. [ABSTRACT FROM AUTHOR]

Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corroded region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA − which suffered the least corrosion −retained some Al in the corroded region. The results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment. [ABSTRACT FROM AUTHOR]

Despite the very demanding operational environment in nuclear reactors, there have been relatively few advanced high-performance materials introduced into fission reactors during the past 50 years. Some of the regulatory and operational barriers to the introduction of high performance materials are briefly discussed, and several examples of potential improvement in current and planned fission reactor systems that could be enabled by advanced structural materials for in-core applications are outlined. Enhanced non-proprietary public-private research and development on advanced structural materials could yield numerous performance, economic, environmental and safety benefits for current boiling water and pressurized water reactors as well as future Generation IV reactor systems such as sodium cooled fast reactors or very high temperature gas cooled reactors. [ABSTRACT FROM AUTHOR]

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300–500 μm thick cladding. [ABSTRACT FROM AUTHOR]

The most direct means of improving the ability of water cooled reactors to withstand excessive cladding oxidation during a loss of coolant accident is to either modify or replace zirconium cladding. It is important to understand what level of agreement is to be expected as a function of systematic differences in steam oxidation testing techniques and instrumentation among testing facilities. The present study was designed to assess the sensitivities of some of the current and proposed reactor cladding materials. Steam oxidation sensitivity of Zircaloy-2, FeCrAl and Mo to O 2 impurities in steam were examined. It was shown that the effect of O 2 impurities is negligible for the two former materials while significant in the case of Mo. [ABSTRACT FROM AUTHOR]

Abstracts Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. The most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Therefore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures. [ABSTRACT FROM AUTHOR]

Assessment of the thermodynamics of SiC corrosion under light water reactor coolant environments suggests that silica formation is always expected in the range of applicable pH and potential. Autoclave testing of SiC-based materials in the absence of ionizing radiation was performed. The kinetics data from these tests, when compared with kinetics of silica dissolution in water and post-exposure characterization of SiC samples, suggest that oxidation of SiC to form silica is the rate-limiting step for recession of SiC in high temperature water. Oxygen activity in water was determined to play an important role in SiC recession kinetics. A simplified model of a power loop shows the effect of silica dissolution from the hot region (resembling fuel) and deposition in the cold regions. [ABSTRACT FROM AUTHOR]

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies. [ABSTRACT FROM AUTHOR]

High-temperature oxidation behavior of SiC coating layer of TRISO fuel particles in 1500–1700 °C steam at 1 atm has been examined inside a zirconia furnace. The SiC coating layers experienced a thickness loss of less than 2.5 μm under these conditions up to 24 h. The thickness of the oxide layer formed under these conditions was consistent with prior steam oxidation tests on high-purity bulk SiC. Upon reducing the presence volatile impurities from the test environment (particularly Al) by conducting the tests inside a zirconia furnace, melting of the silica layer at 1700 °C was avoided. [ABSTRACT FROM AUTHOR]

The consolidation mechanism and resulting properties of the silicon carbide (SiC) matrix of fully ceramic microencapsulated (FCM) fuel form are discussed. The matrix is produced via the nano-infiltration transient eutectic-forming (NITE) process. Coefficient of thermal expansion, thermal conductivity, and strength characteristics of this SiC matrix have been characterized in the unirradiated state. An ad hoc methodology for estimation of thermal conductivity of the neutron-irradiated NITE–SiC matrix is also provided to aid fuel performance modeling efforts specific to this concept. Finally, specific processing methods developed for production of an optimal and reliable fuel form using this process are summarized. These various sections collectively report the progress made to date on production of optimal FCM fuel form to enable its application in light water and advanced reactors. [ABSTRACT FROM AUTHOR]

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235 U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (∼0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239 Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235 U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding. [ABSTRACT FROM AUTHOR]

Growth and microstructure of a protective or nonprotective SiO2 scale and the subsequent volatilization of scale formed on high-purity chemical vapor deposited ( CVD) SiC and nuclear-grade SiC/ SiC composites have been studied during high-temperature 100% steam exposure. The environmental parameters of interest were temperature from 1200°C to 1700°C, pressure of 0.1 to 2 MPa and flow velocities of 0.23 to 145 cm/s. Scale microstructure was characterized via electron microscopy and X-ray diffractometry. The Arrhenius dependence of the parabolic oxidation and linear volatilization rate constants were determined. The linear volatilization rate exhibited a strong dependence on steam partial pressure with a weaker dependence on flow velocity. At high steam pressures, the oxide scale developed substantial porosity, which significantly accelerated material recession. The dominant oxide phase for the conditions studied was cristobalite. The oxidation behavior of SiC/ SiC composite was strongly dependent on the state of the surface, specifically whether steam could find easy entry into the material via surface-exposed interface layers. For the case where these as-machined interfaces were surface coated with matrix CVD SiC, composite recession was found to be essentially that of high-purity CVD SiC. [ABSTRACT FROM AUTHOR]

Abstract: The effect of variation in thermal conductivity of light water reactor fuel elements on core response during loss-of-coolant accident scenarios is examined. Initially, a simplified numerical analysis is utilized to determine the time scales associated with dissipation of stored energy from the fuel into the coolant once the fission reaction is stopped. The analysis is then followed by full reactor system thermal–hydraulics analysis of a typical boiling and pressurized water reactor subjected to a large break loss-of-coolant accident scenario using the TRACE code. Accordingly, sensitivity analyses to examine the effect of an increase in fuel thermal conductivity, up to 500%, on fuel temperature evolution during these transients are performed. Given the major differences in thermal–hydraulics design aspects of boiling and pressurized water reactors, different fuel and temperature responses during the simulated loss-of-coolant transients are observed. [Copyright &y& Elsevier]

Abstract: The feasibility of using SiCf–SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components. [Copyright &y& Elsevier]

Abstract: The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms. [Copyright &y& Elsevier]

Abstract: Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics. [Copyright &y& Elsevier]

Abstract: This paper addresses certain key feasibility issues facing the application of SiC-matrix microencapsulated fuels for light water reactor application. Issues addressed are the irradiation stability of the SiC-based nano-powder ceramic matrix under LWR-relevant irradiation conditions, the presence or extent of reaction of the SiC matrix with zirconium-based cladding, the stability of the inner and outer pyrolytic graphite layers of the TRISO coating system at this uncharacteristically low irradiation temperature, and the state of the particle–matrix interface following irradiation which could possibly affect thermal transport. In the process of determining these feasibility issues microstructural evolution and change in dimension and thermal conductivity was studied. As a general finding the SiC matrix was found to be quite stable with behavior similar to that of CVD SiC. In magnitude the irradiation-induced swelling of the matrix material was slightly higher and irradiation-degraded thermal conductivity was slightly lower as compared to CVD SiC. No significant reaction of this SiC-based nano-powder ceramic matrix material with Zircaloy was observed. Irradiation of the sample in the 320–360°C range to a maximum dose of 7.7×1025 n/m2 (E >0.1MeV) did not have significant negative impact on the constituent layers of the TRISO coating system. At the highest dose studied, layer structure and interface integrity remained essentially unchanged with good apparent thermal transport through the microsphere to the surrounding matrix. [Copyright &y& Elsevier]

Abstract: The kinetics of hydrogen desorption from zirconium hydride is important in many nuclear design and safety applications. In this paper, a coordinated experimental and modeling study has been used to explicitly demonstrate the applicability of existing kinetic theories for hydrogen desorption from zirconium hydride and α-zirconium. A static synthesis method was used to produce δ-zirconium hydride, and the crystallographic phases of the zirconium hydride were confirmed by X-ray diffraction (XRD). Three obvious stages, involving δ-zirconium hydride, a two-phase region, and α-zirconium, were observed in the hydrogen desorption spectra of two zirconium hydride specimens with H/Zr ratios of 1.62 and 1.64, respectively, which were obtained using thermal desorption spectroscopy (TDS). A continuous, one-dimensional, two-phase moving boundary model, coupled with the zero- and second-order kinetics of hydrogen desorption from δ-zirconium hydride and α-zirconium, respectively, has been developed to reproduce the TDS experimental results. A comparison of the modeling predictions with the experimental results indicates that a zero-order kinetic model is valid for description of hydrogen flux away from the δ-hydride phase, and that a second-order kinetic model works well for hydrogen desorption from α-Zr if the activation energy of desorption is optimized to be 70% of the value reported in the literature. [Copyright &y& Elsevier]

Alternative fuel cladding materials to Zr alloys are being investigated for enhanced accident tolerance, which specifically involves oxidation resistance to steam or steam–H2 environments at ⩾1200°C for short times. Based on a comparison of a range of commercial and model alloys, conventional austenitic steels do not have sufficient oxidation resistance with only ∼18Cr–10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application. Results at 1350°C indicated that FeCrAl alloys and CVD SiC remain oxidation resistant in steam. At 1200°C, high (⩾25% Cr) ferritic alloys appear to be good candidates for this application. Higher pressures (up to 20.7bar) and H2 additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys, but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed for type 317L tubing in a H2–50%H2O environment at 10.3bar compared to 100% H2O. [ABSTRACT FROM AUTHOR]

Abstract: The viability of advanced oxidation-resistant Fe-base alloys to protect zirconium from rapid oxidation in high-temperature steam environments has been examined. Specimens were produced such that outer layers of FeCrAl ferritic alloy and Type 310 austenitic stainless steel were incorporated on the surface of zirconium metal slugs. The specimens were exposed to high-temperature 0.34MPa steam at 1200 and 1300°C. The primary degradation mechanism for the protective layer was interdiffusion with the zirconium, as opposed to high-temperature oxidation in steam. The FeCrAl layer experienced less degradation and protected the zirconium at 1300°C for 8h. Constituents of the Fe-base alloys rapidly diffused into the zirconium and resulted in the formation of various intermetallic layers at the interface and precipitates inside the bulk zirconium. The nature of this interaction for FeCrAl and 310SS has been characterized by use of microscopic techniques as well as computational thermodynamics. Finally, a reactor physics discussion on the applicability of these protective layers in light-water-reactor nuclear fuel structures is offered. [Copyright &y& Elsevier]

Abstract: The extent and nature of the chemical interaction between the outermost coating layer of coated fuel particles embedded in zirconium metal during fabrication of metal matrix microencapsulated fuels were examined. Various particles with outermost coating layers of pyrocarbon, SiC, and ZrC have been investigated in this study. ZrC–Zr interaction was the least substantial, while the PyC–Zr reaction can be exploited to produce a ZrC layer at the interface in an in situ manner. The thickness of the ZrC layer in the latter case can be controlled by adjusting the time and temperature during processing. The kinetics of ZrC layer growth is significantly faster from what is predicted using literature carbon diffusivity data in ZrC. SiC–Zr interaction is more complex and results in formation of various chemical phases in a layered aggregate morphology at the interface. [Copyright &y& Elsevier]

Abstract: The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms. [Copyright &y& Elsevier]

Abstract: The metal matrix microencapsulated (M3) fuel concept for light water reactors (LWRs), consisting of coated fuel particles dispersed in a zirconium metal matrix, is introduced. Fabrication of M3 fuels by hot pressing, hot isostatic pressing, or extrusion methodologies has been demonstrated over the temperature range 800–1050°C. Various types of coated fuel particles with outermost layers of pyrocarbon, SiC, ZrC, and TiN have been incorporated into the zirconium metal matrix. Mechanical particle–particle and chemical particle–matrix interactions have been observed during the preliminary characterization of as-fabricated M3 specimens. Irradiation of three M3 rodlets with surrogate coated fuel particles was carried out at mean rod temperature of 400°C to 4.6dpa in the zirconium metal matrix. Due to absence of texture in the metal matrix no irradiation growth strain (<0.09%) was detected during the post-irradiation examination. [Copyright &y& Elsevier]

Abstract: The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina–yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder–fuel particle mixture at a temperature of 1800–1900°C using compaction pressures of 10–20MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle–matrix interface. [Copyright &y& Elsevier]

Abstract: Monophase thorium–zirconium hydrides (ThZr2H x ) have been fabricated starting from a metallic alloy and the hydrogen stoichiometry determined by X-ray diffraction. Incoherent Quasielastic Neutron Scattering (IQNS) on the hydrides was conducted over the temperature range 650–750K at the Backscattering Silicon Spectrometer (BASIS) at the Spallation Neutron Source (SNS) at ORNL. The isotropic Chudley–Elliott model was utilized to analyze the quasielastic linewidth broadening data as function of momentum transfer. The diffusion coefficient and average jump distance of hydrogen atoms in ThZr2H5.6 and ThZr2H6.2 were extracted from the measurements. [Copyright &y& Elsevier]

Abstract: Three sets of independent experiments were conducted to determine the kinetics of hydrogen desorption from and adsorption on δ-zirconium hydrides. One method involved measurement of hydrogen gas pressure-buildup as a result of dehydriding in a closed vessel. The other two involved thermogravimetric experiments, measuring the rate of mass loss during dehydriding under vacuum. Zeroth-order desorption and first order (with respect to gas-phase hydrogen) adsorption kinetics was determined. The Arrhenius dependence of rate constants showed excellent agreement among different experimental data. The activation energies for desorption and adsorption processes were determined as 205±8 and 86±15kJmol−1 respectively. [Copyright &y& Elsevier]

We report the case of a 53-year-old man with solar urticaria ( SU) not responding to histamine antagonist therapy. Using intradermal testing with the patient's irradiated serum, we demonstrated that his SU was elicited by visible light only. This type of SU is uncommon, and makes preventive measures such as sun block largely ineffective. However, the patient had an extraordinary response to the treatment with omalizumab. This case report highlights the diagnostic process and the success of omalizumab for the treatment of SU after 16 months of follow-up. [ABSTRACT FROM AUTHOR]

Abstract: The transient response of a uranium zirconium hydride (U0.31ZrH1.6) fuel element to conditions typical of a light water reactor reactivity insertion accident has been studied. Hydrogen diffusion within the fuel is treated as a function of the time-dependent temperature profile. Temperature and hydrogen concentration dependence of thermal properties of the fuel is also considered. The set of coupled equations describing the transient conduction of heat and diffusion of hydrogen are solved with a Crank–Nicolson discretization scheme for heat diffusion and an explicit Euler scheme for hydrogen diffusion. [Copyright &y& Elsevier]

Abstract: Two uranium–thorium–zirconium hydrides, (UTh4Zr10)H1.9 and (U4Th2Zr9)H1.5, have been fabricated and characterized. Fabrication involved arc melting of the constituent pure metals to form homogenous alloys, followed by hydriding at elevated temperatures in a hydrogen gas environment. The compounds were characterized by X-ray powder diffractometry as well as scanning and transmission electron microscopy. These methods revealed a multi-phase mixture of δ-zirconium hydride (ZrH1.6+ x ), thorium–zirconium hydride (ThZr2H7− x ), and uranium metal. The elastic modulus was mapped across the microstructure using nanoscale dynamic stiffness mapping. The elastic modulus of ThZr2H7− x phase is found to be 172GPa. [Copyright &y& Elsevier]

Neutron irradiation tests were carried out on 3D-printed SiC derived from binderjet additive manufacturing and chemical vapor infiltration. Irradiation was carried to 2.3 dpa over a temperature range of 400–850 °C. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. No degradation in strength was observed in the material before or after irradiation, at various temperatures, or in different orientations. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed. [ABSTRACT FROM AUTHOR]

An integrated safety design and radionuclide (RN) retention strategy is developed to support the Transformational Challenge Reactor (TCR) demonstration. This demonstration aims to showcase viability for rapid deployment of a novel reactor by leveraging the advances in materials, manufacturing, and computational sciences through a highly integrated and agile design and development approach. This strategy provides a logical description and understanding of how RNs are contained within the facility. Rather than discussing fission product barriers individually between separate design and safety basis reports, this paper provides a consistent description and narrative to better facilitate regulatory interactions and focus safety design efforts. The principal barriers credited include the various coating layers in the tristructural isotropic (TRISO) fuel particle, the silicon carbide (SiC) matrix hosting the particles within the fuel element, the helium pressure boundary, and the confinement system. The choice and assumed performance of the credited barriers are highly conservative, which is a direct reflection of the low hazard that the TCR demonstration presents and the need to simplify and focus the safety review process accordingly. However, the strategy and the associated framework are generalized and may be adopted and tailored to support other advanced reactor demonstration efforts. [ABSTRACT FROM AUTHOR]

• A summary compilation of the properties of non-irradiated YH x was reported. • The correlation between the thermomechanical properties and the hydrogen concentration was revealed. • Microstructural characterization of YH x indicates that the grain structure is highly dependent on the fabrication process. Yttrium hydride is an optimal choice for a high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. However, a complete database of the thermomechanical properties of yttrium hydride is not available yet, although it is much needed to understand and predict the moderator performance during service in reactors. In this paper, we report the properties of unirradiated bulk yttrium hydride as a function of hydrogen concentration—including density, crystal structure, specific heat capacity, thermal diffusivity, thermal conductivity, hardness, elastic/shear moduli, Poisson's ratio, fracture strength, microstructure, and thermal stability—providing a baseline measurement for the subsequent neutron irradiation response study of yttrium hydride. The recommended empirical treatment of the data is suggested. In addition, other properties (i.e., hydrogen retention, thermal hydrogen migration, and irradiation response) that needs to be investigated are discussed. [ABSTRACT FROM AUTHOR]

The fuel form developed for the Transformational Challenge Reactor demonstration program leverages recent advances in manufacturing, materials, and computational sciences, delivering a new architecture for production of high-performance microencapsulated nuclear fuels. The fuel consists of conventionally manufactured uranium nitride tristructural isotropic fuel particles embedded inside a 3D-printed silicon carbide matrix. This paper describes the overall architecture and manufacturing process for this fuel form, its properties and behavior, and the ongoing development activities. [ABSTRACT FROM AUTHOR]

Additive manufacturing (AM) is a rapidly evolving technology being considered for nuclear applications. A special focus on AM to fabricate nuclear-grade silicon carbide (SiC) is explored in this paper. First, we present currently available AM processing options for SiC. AM methods commonly used for other ceramics, in which the feedstocks are forms of polymers, powders, and/or reactive chemical vapors, are also applicable to SiC. SiC phases are formed by pyrolysis of pre-ceramic polymer, direct reaction of powder precursors, sintering of SiC powders, or chemical vapor deposition/infiltration. Second, we discuss how the different microstructures of SiC materials fabricated by various processing methods affect their behavior in nuclear environments. Third, we discuss state-of-the-art AM technologies for the fabrication of relatively pure SiC, which show great potential to retain its strength under neutron irradiation: (1) binder jet printing followed by chemical vapor infiltration, (2) laser chemical vapor deposition, and (3) selective laser sintering of SiC powders. [ABSTRACT FROM AUTHOR]