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Objective: Funic presentation is a risk factor for umbilical cord prolapse. In the majority of cases, umbilical cord prolapse is an obstetric emergency. Currently there is no strategy to prevent umbilical cord prolapse in patients with funic presentation. Placing a third trimester cervical cerclage was proposed, as a mechanical barrier of prolapse, in women with funic presentation. Materials and Methods: Twelve patients with funic presentation, detected between 23 and 34 weeks of gestation, were included in the study. Patients were informed that although cervical cerclage have been widely used in obstetrics, it had not been offered in patients with funic presentation. Gestational age at delivery, mode of delivery, location of umbilical cord at birth, presence or absence of umbilical cord prolapsed, and neonatal condition at birth were recorded in all cases. Results: Patients’ age varied from 21 to 40 years, mean of 28 ± 6.2 years, and gestational age at the time of diagnosis of funic presentation varied from 26 to 34 weeks of pregnancy. All patients were seen weekly after the cervical cerclage placement to assess the umbilical cord location. Seven patients had vertex presentation, four had breech presentation, and one had transverse lie presentation. A follow-up sonogram of the umbilical cord revealed the following: funic presentation was persistent throughout the entire pregnancy in five patients, disappeared in four, and became intermittent in three. The delivery mode was as follows: cesarean section was performed in eight patients (five with persistent funic presentation, two with intermittent, and one with resolved funic presentation because of failure to progress in labor). There were five patients who delivered vaginally. Funic presentation was confirmed in all patients who underwent cesarean sections. None of the patients had had an umbilical cord prolapse. Conclusion: In view of high perinatal mortality of umbilical cord prolapses, placement of cervical cerclage could be added to obstetric armamentarium in patients with funic presentation, after proper counseling.

Zusammenfassung. Vaskulitiden sind eine heterogene Gruppe von Erkrankungen, die unterschiedlich klassifiziert werden, zum Beispiel nach der Grösse des betroffenen Gefässes oder nach primärer und sekundärer Ätiologie. Die Haut ist am häufigsten betroffen ist, sie kann sowohl isoliert als auch im Rahmen von systemischen Formen involviert sein. Die Kombination der Hautläsionen, deren anatomischer Lokalisation und Informationen zum Zeitablauf geben Hinweise für die Differenzialdiagnose. Purpura, Blasen, Nekrosen, Ulzerationen und allenfalls eine Livedo sind charakteristische Manifestationen. Konstitutionelle Symptome wie Gewichtsverlust, Abgeschlagenheit, Fieber und Arthralgien sind hinweisend auf eine systemische Form. Wichtig ist die differenzialdiagnostische Abgrenzung von Vaskulopathien, die sich ähnlich manifestieren können. Am häufigsten beim Erwachsenen ist die kutane leukozytoklastische Angiitis, beim Kind die IgA-Vaskulitis (Purpura Schönlein-Henoch). Bei diesen kommen diverse Auslöser in Frage: Infektionen, Medikamente, Autoimmunerkrankungen und Malignome, wobei bis zu 50% ätiologisch ungeklärt bleiben. Hautbiopsien und Laborparameter, allenfalls ergänzt durch Bildgebung, sind wichtige Schritte im Abklärungsprozedere. Die Behandlung richtet sich primär auf die Elimination einer möglichen Ursache. Die idiopathische kutane Form bildet sich meistens spontan zurück, die Behandlung ist symptomatisch. Bei schwererem Verlauf werden primär topisch Kortikosteroide oder Calcineurinantagonisten angewandt. Bei Therapieresistenz werden systemische Immunsuppressiva eingesetzt.

This work introduces a meshed debonding model intended for use with tristructural isotropic (TRISO) fuel particles. The ability to better examine the potential effects of interface debonding betwee...

The advance in technology allows for the development of different CT scanners in the field of dual-energy computed tomography (DECT). In particular, a recently developed detector-based technology can collect data from different energy levels, thanks to its layers. The use of this system is suited for material decomposition with perfect spatial and temporal registration. Thanks to post-processing techniques, these scanners can generate conventional, material decomposition (including virtual non-contrast (VNC), iodine maps, Z-effective imaging, and uric acid pair images) and virtual monoenergetic images (VMIs). In recent years, different studies have been published regarding the use of DECT in clinical practice. On these bases, considering that different papers have been published using the DECT technology, a review regarding its clinical application can be useful. We focused on the usefulness of DECT technology in gastrointestinal imaging, where DECT plays an important role.

Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and subcomponents can be manufactured in parallel over time frames of weeks rather than months or years. To meet key program initiatives (e.g., timeline, material use), several constraints—including fissile material availability (less than 250 kg of HALEU), component availabilities, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small (less than one cubic meter in volume) cores with near-term viability. The fast-spectrum designs did not meet the fissile material constraint, so the thermal-spectrum systems became the primary design focus. Since significant progress has been made on advanced moderator materials (YHx) under the TCR program, gas-cooled thermal-spectrum systems using less than 250 kg of HALEU that occupy less than 1 m3 are now feasible. The designs for two of these systems have been evolved and matured. In both thermal-spectrum design concepts, bidirectional coolant flow is used. Coolant flows down through YHx moderator elements and is reversed in a bottom manifold and core support structure, and then flows up though or around the fuel elements. The main difference between the two thermal-spectrum design concepts is the fuel elements—one uses traditional UO2 ceramic fuel, and the other uses UN-bearing TRISO fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems are assessed and summarized herein.

Objective: Umbilical cord prolapse is a rare event complicating 0.17%–0.62% of all pregnancies. Funic presentation is a known risk factor for umbilical cord prolapse. Currently, there is no strategy to prevent umbilical cord prolapse in patients with funic presentation. The novel technique used is placement of late cervical cerclage to create a mechanical barrier and prevent an umbilical cord prolapse. Methods: Six patients with a sonographically detected funic presentation were included in the study. Funic presentation was defined as the sonographic presence of the umbilical cord below the presenting part using both transabdominal and transvaginal sonography. Cord prolapse was defined as an umbilical cord seen or palpated below the presenting part. Cervical cerclage was placed in patients with persistent funic presentations, which is the detection of the umbilical cord below the presenting part, on two or more sonograms at least a week apart. Results: Cervical cerclages were placed in six patients with funic presentation between 28 and 34 weeks of gestation without immediate complications. All patients were delivered by a cesarean section between 35 and 38 weeks of pregnancy. None experienced umbilical cord prolapse. Funic presentation was confirmed at birth in all cases. Apgar scores varied between 7 and 10. Conclusion: It appears that cervical cerclage may be an effective measure to prevent umbilical cord prolapse in cases of known persistent funic presentation.

Characterizing oxide nuclear fuels is difficult due to complex fission products, which result from time-evolving system chemistry and extreme operating environments. Here, we report a machine learning-enhanced approach that accelerates the characterization of spent nuclear fuels and improves the accuracy of identifying nanophase fission products and bubbles. We apply this approach to commercial, high-burnup, irradiated light-water reactor fuels, demonstrating relationships between fission product precipitates and gases. We also gain understanding of the fission versus decay pathways of precipitates across the radius of a fuel pellet. An algorithm is provided for quantifying the chemical segregation of the fission products with respect to the high-burnup structure, which enhances our ability to process large amounts of microscopy data, including approaching the atomistic-scale. This may provide a faster route for achieving physics-based fuel performance modeling.

Ceramic-metallic nuclear fuels are a candidate fuel for nuclear thermal propulsion systems due to their high heat transport properties, which are necessary in very high-temperature environments. Th...

Nano-oxide precipitates in a modern nanostructured ferritic alloy were investigated after extreme thermomechanical processing into a thin-walled tube geometry. It was found that the morphology of the precipitates changed from spherical to rod-shaped, with some increasing to aspect ratios of up to 9, despite the precipitate volume fraction (0.3%) and number density (> 1023 m−3) of precipitates remaining unchanged. High-resolution electron microscopy showed that the precipitates likely remained coherent with the Fe-matrix, while atom probe tomography confirmed that the precipitate compositions remained unaffected by the transformation. The morphological change was attributed to the shearable nature of the (Y,Ti,O)-rich precipitates, indicating they should be considered as “soft” obstacles to dislocation motion. The elongation was most pronounced in larger (>5 nm) precipitates, which may be caused by preferential dissolution of the smallest (1–3 nm) precipitates followed by the competition between re-precipitation and solute diffusion to larger precipitates during recovery heat treatments.

Background: Mycoplasma pneumoniae pneumonia is sometimes associated with skin or mucous membrane eruptions. Available reviews do not address the association of Chlamydophila pneumoniae pneumonia with skin eruptions. We therefore conducted a systematic review of the literature addressing this issue. The National Library of Medicine, Excerpta Medica, and Web of Science databases were employed. Summary: In two reports, skin lesions and especially urticaria were more common (p < 0.05) in atypical pneumonia caused by C. pneumoniae as compared with M. pneumoniae. We found 47 patients (n = 16; ≥18 years, n = 31) affected by a C. pneumoniae atypical pneumonia, which was associated with erythema nodosum, erythema multiforme minus, erythema multiforme majus, isolated mucositis, or cutaneous vasculitis. We also found the case of a boy with C. pneumoniae pneumonia and acute generalized exanthematous pustulosis. We did not find any case of C. pneumoniae respiratory infection associated with either Gianotti-Crosti syndrome, pityriasis lichenoides et varioliformis acuta Mucha-Habermann, or varicella-like skin eruptions.

Background Percutaneous tracheostomy is frequently performed in intensive care units in patients who require prolonged mechanical ventilation. The first crucial step for the physician in these procedures is the precise needle insertion into the trachea. The primary aim of this technical note was to test the new filament and share our experiences in the implementation of the new device. The secondary aim was to show how a physician with basic training in computer-aided design and three-dimensional (3D) printing could independently create useful devices for clinical practice. Methods To simplify this referred clinical procedure and increase its safety, 3D printing and a new medical filament were used to develop a new translaryngeal Tracheostomy Needle Introducer (tTNI) for use in conjunction with the Fantoni’s method of percutaneous tracheostomy. The tTNI is composed of three parts: a support to fit on the rigid endotracheal tube of the Fantoni kit, an external particular shaped arm, and an introducer for the needle. The latest version of the device used a new filament based on a polyester matrix certified for skin contact that was sterilizable in a standard autoclave. Post-printing, minor technical interventions were required to correct small material deformities. Conclusions Our experiences with the thread and the technical features of the material were reported herein in conjunction with some suggestions on how to solve the most frequently encountered problems. The 3D printing technique allows physicians to directly manage the prototyping process of new medical devices, making this process completely independent. The speed of the prototyping process and the testing of each piece allow faster creation of a prototype than with traditional industrial methods. Finally, the new biomedical filaments offer endless possibilities of creation and modelling.

Steady-state internal flow boiling experiments were conducted on various materials, including accident tolerant fuel cladding material Fe–12Cr–6Al (C26M2) alloy, Zircaloy, and metal-based m...

The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 × 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO2/Zircaloy with UO2/FeCrAl would require an average enrichment increase of 0.6% 235U throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO2/FeCrAl models with these geometric/enrichment specifications matched the base UO2/Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel centerline temperature, comparable axial elongation and radial displacement, and longer time to gap closure compared to a Zircaloy-clad fuel rod. This fuel performance analysis was primarily based on the commercial Kanthal APMT FeCrAl alloy but also used available data for the C35M FeCrAl alloy developed at Oak Ridge National Laboratory.

The successful development of advanced materials such as nanostructured ferritic alloys (NFA) for next generation nuclear reactor concepts and ultra-supercritical steam power plants requires information on long term thermal creep. To support this initiative, this work examines the NFA MA957 (Fe-14Cr-1.0Ti-0.3Mo + 0.3Y2O3 in wt%) crept to 61,251 h at 825 °C and 70 MPa, the longest creep test available to date for this material. Using atom probe tomography and electron microscopy, it is shown that the grain size and nanoprecipitate size/composition are unaffected following this 7 yr creep test, although significant porosity is noted throughout the microstructure attributed to microvoid coalescence and growth.

The initial results of a post-irradiation examination study conducted on a SiC-SiC tubular specimen irradiated under a high radial heat flux are presented herein. The elastic properties of the specimen were evaluated before and after the irradiation using the resonant ultrasound spectroscopy (RUS) technique. The composite tubular specimen was considered as an orthotropic elastic with nine elastic constants (Young's moduli, shear moduli and Poisson's ratios—three components of each) for representing its full elastic deformation behavior. All the elastic moduli decreased after irradiation; the reduction was as high as 35% in one of the moduli. The significant decrease in the moduli indicates the presence of microcracks. The results from a computational study show significant stress development in the specimen due to irradiation, primarily caused by differential swelling across the thickness of the specimen. The evaluated stresses exceed the proportional limit stress of the material, indicating the likelihood of matrix microcracking, and thus corroborating the results obtained from RUS. X-ray Computed Tomography (XCT) study confirmed the presence of cracks in the irradiated specimen. These cracks occurred at the inner region of the specimen and propagated in axial and hoop directions. These XCT results are in agreement with the RUS results and stress distribution results from the computational study.

Fast reactor fuel cladding candidate materials require proficiency in extreme environments consisting of high temperatures and irradiation doses in excess of 150 displacements per atom (dpa). Nanostructured oxide dispersion strengthened (ODS) alloys have been developed extensively for this purpose due to their notable high temperature strength, creep resistance, and irradiation resistance. However, their properties can deteriorate if interstitial impurities such as C and N are not well controlled during the fabrication process. A new Fe-12Cr nanostructured ODS alloy OFRAC ( O ak Ridge F ast R eactor A dvanced Fuel C ladding) with solute additions of Mo, Ti, and Nb has been developed to provide the desired properties mentioned above while simultaneously sequestering impurities within the matrix. After extrusion at 850 °C, the as-extruded microstructure consists of an average 490 nm grain size and a high number density (6.8 × 1023 m-3) of 2.2 nm diameter (Y,Ti,O) nanoclusters distributed homogeneously in the matrix. Atom probe tomography investigations suggest non-stochiometric compositions for the smallest nanoclusters. In addition, a second population of nanometer scale (Nb,Ti) rich carbonitrides is also present in the microstructure that captures the potentially detrimental C and N impurity atoms present in the matrix. Atom probe tomography results indicate elemental segregation of Cr, Mo, and Nb to grain boundaries in the as-extruded material, consistent with previous investigations of solid solution strengthening by solute additions. The ability of OFRAC to sequester impurities introduced from the powder metallurgical approach to nanostructured ferritic alloy development, compounded with its beneficial mechanical properties, makes this alloy a competitive candidate for fast reactor applications. This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ).

The kernel of fully ceramic microencapsulated (FCM) fuel requires a material with high fissile density. For that reason, among others, UN is the appropriate chemical state for low enriched U fuel. The UN kernels are spheres ∼800 μm in diameter made using a sol-gel process. The effect of additives on the chemistry and density of the UN microspheres are investigated in this work. Gadolinium nitrate hexahydrate, Gd2O3, B, SiO2 and SiC were incorporated into the sol-gel broth in varying concentrations. It was found that Gd can serve as both a sintering aid and burnable poison when added to the sol-gel broth as a Gd(NO3)36H2O. However, even with the increased theoretical density of the UN microspheres, the U density was still too low for the FCM design that replaces UO2 pellets in a commercial light water reactor. Silicon carbide and B were also successfully added but produced a lower density final product. Other sol-gel processing variables were investigated. The pour density of the sol-gel feedstock was found to influence the final converted UN kernel density.

Fabrication of monolithic UO2 and UO2-Mo composites from Mo metal powder feedstock and UO3 spheres derived from internal gelation was explored. The addition of 10 vol% Mo as a secondary phase to UO2 led to an increase of up to 30% in thermal conductivity compared to syntheses in which only pure UO2 was used. Using coarse and fine Mo metal powder feedstock led to different microstructures, influencing the thermal conductivity of the resulting composites.

An advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the “driver tube” concept and allows for high-speed testing with internal pressure impulses of 10–1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360° view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

Steady and transient (with a heating rate of 685 °C/s) internal-flow CHF (Critical Heat Flux) experiments were conducted under atmospheric pressure at a fixed inlet temperature (40 °C or 60 °C) and mass flow (300 kg/m2 s) on Fe-13Cr-6Al alloy, Inconel 600 and Zircaloy-4 tube samples. Multiple experiments were repeated on the same specimen to investigate the effect of surface characteristic changes (i.e., roughness, wettability, and oxide scale morphology) on the occurrence of CHF. Despite notable changes of wettability, roughness, and oxide layer characteristics on samples that had already been subjected to CHF, measured flow CHF remained unchanged throughout repeated experiments for tested materials. This demonstrates that the surface effects on flow CHF are limited in the test conditions. In the steady-state flow boiling condition, Fe-13Cr-6Al alloy demonstrated a 22% and 14% increase in CHF compared to Zircaloy-4 and Inconel 600, respectively. Compared to the 2006 Groeneveld CHF lookup table, Fe-13Cr-6Al alloy gives a 13% increase in the tested flow boiling condition. Material properties are considered primarily responsible for the observed CHF differences among the tested materials. The surface thermal economy parameter ( ρ c p 3 / 2 k ) is proposed as an explanation for the observed CHF differences; this parameter is related to material’s ability to avoid an irreversible dry spot formation. The apparent disagreement of Zircaloy-4 CHF with both the look up table predictions and Inconel 600 shows the limitation of departure of nucleate boiling (DNB) evaluations that do not consider cladding materials. The transient Fe-13Cr-6Al CHF is 39% and 23% higher than the lookup table prediction and the steady-state condition experimental result, respectively.

We conducted reactor performance calculations to assess the potential design basis accident performance of HTGR fuel designs. Three Fully Ceramic Microencapsulated (FCM) fueled HTGR designs were developed in a previous work (Lu et al., 2018). The maximum fuel temperature in the cores fueled by these three FCM fuels was predicted to be higher than that in the reference 350-MWt mHTGR core in both normal operating conditions and during representative design basis accidents (Lu and Brown, 2019). To better understand the potential safety margins in mHTGR design basis accidents, we performed thermal-hydraulics sensitivity studies to investigate how maximum fuel temperature varies considering various parameters, e.g. thermal properties, within the ranges corresponding to the differences between the FCM-fueled prismatic mHTGR cores and the reference core with conventional fuel compacts. We found that the difference in the steady-state axial power distribution contributed the most to the difference in the maximum fuel temperature, in both normal operation and design basis accidents. Experimental data suggested that the annealing process of irradiation defects in SiC would be rapid at mHTGR relevant fuel temperatures. The bounding potential impact of the SiC annealing on the maximum fuel temperature was analyzed considering both the thermal conductivity recovery and the Wigner energy release due to the annealing of SiC. We found that the defect annealing process in SiC would at most increase the peak maximum fuel temperature of an FCM-fueled core by 40 K in loss of forced cooling accidents and by 10 K in a control rod withdrawal accident. Additional experiments on the SiC defect annealing kinetics and Wigner energy release in more relevant conditions are needed.

Oxide dispersion strengthened (ODS) FeCrAl alloy (Fe-12Cr-5Al wt%) was neutron irradiated to 1.8 displacements per atom (dpa) at 215, 357, and 557 °C to investigate (Y,Al,O) nano-oxide stability and Cr-rich α′ precipitate formation using atom probe tomography. The nano-oxide sizes remain unchanged after irradiation, but the compositions are dependent on irradiation temperature due to competition between ballistic dissolution and precipitate reformation. Cr-rich clusters observed after 357 °C irradiation have lower Cr contents (61 at.%) than equilibrium α′, with a lower number density than wrought FeCrAl alloys. This suggests a suppression of α′ precipitation due to the high sink strength of ODS FeCrAl.

Three distinct composite architectures of silicon carbide (SiC) and pyrocarbon (PyC) were incorporated into the SiC coating layer of tristructural-isotropic (TRISO) nuclear fuel particles. The composite architectures are meant to increase the resistance of SiC coating layer to cracking and fission product attack during operation and accident scenarios. All composite layers were produced using the existing fluidized bed chemical vapor deposition apparatus that is used for production of TRISO fuel particles without modifications. Detailed characterization of the composite microstructure was carried out via optical and electron microscopy. Nano-indentation examination confirms that mechanical properties of the SiC phase was not affected in the composite architectures, however, the resistance to crack propagation in this coating layer was greatly increased in all cases when compared to the reference monolithic coating layer. The stress required to debond the SiC-inner PyC interface in the reference TRISO particles was determined to be ∼1 GPa using micropillar compression technique. The high strength may explain the ease of crack propagation from the inner PyC to SiC in the reference design. In the composite architectures, the means of crack deflection were effectively incorporated at this interface. Finite element analysis of stress evolution in the fuel particles during normal operation with the reference and composite SiC coating layer architectures did not show any significant differences between the variants.

The purpose of this investigation is to experimentally quantify the stored energy release upon thermal annealing of previously irradiated high-purity silicon carbide (SiC.) Samples of highly-faulted polycrystalline CVD β-SiC and single crystal 6H SiC were irradiated in a mixed spectrum fission reactor near 60 °C in a fluence range from 5 × 1023 to 2 × 1026 n/m2 (E > 0.1 MeV), or about 0.05–20 dpa, in order to quantify the stored energy release and correlate the release to the observed microscopic swelling, lattice dilation, and microstructure as observed through TEM. Within the fluence of this study the crystalline material was observed to swell to a remarkable extent, achieving 8.13% dilation, and then cross a threshold dose for amorphization at approximately 1 × 1025 n/m2 (E > 0.1 MeV) Once amorphized the material attains an as-amorphized swelling of 11.7% at this irradiation condition. Coincident with the extraordinary swelling obtained for the crystalline SiC, an equally impressive stored energy release of greater than 2500 J/g at the critical threshold for amorphization is inferred. As expected, following amorphization the stored energy in the structure diminishes, measured to be approximately 590 J/g. Generally, the findings of stored energy are consistent with existing theory, though the amount of stored energy given the large observed crystalline strain is remarkable. The overall conclusion of this work finds comparable stored energy in SiC to that of nuclear graphite, and similar to graphite, a stored energy release in excess of its specific heat in some irradiation conditions.

The mechanical response of a nuclear-grade silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite was investigated under mechanical loading conditions mimicking the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA). In a RIA, cladding deformation and failure can be induced by the rapid thermal expansion of the nuclear fuel. A pulse-controlled modified-burst test was used to investigate RIA-like PCMI scenarios on SiC/SiC composite samples at pulse widths from 12 to 100 ms. The strain-driven nature of the cladding sample deformation was due to the rapid internal pressurization and subsequent expansion of a secondary tube. A digital-image correlation technique was used to measure strains from the speckle-painted outer surface of the tubes. The failure strains of samples tested at slower rates, such as RIA event durations of 52 and 100 ms, showed good agreement with the literature-reported values for similar composites tested at slow strain rates. Additionally, the failure strain showed good agreement with reference expansion-due-to-compression tests at slow strain rate. However, a decrease in the failure strain was determined for the fast-rate (12 ms) tests. This indicated that the failure strain of these composites might be influenced by the strain rate during RIA-like events. The failure strains observed in the tests corresponded to local energy depositions of approximately 50 cal/g UO2 from hot zero power, with an initial condition of pellet–cladding gap closure prior to the event. In-pile transient testing of these concepts that would result in hoop strain due to PCMI in the range of 0.5–1.0% is recommended.

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.

Additional file 6. Supplemental_snap.

Additional file 4. Supplemental_exagon.

Additional file 5. Supplemental_TNI arm.

Additional file 7. Supplemental_head.

Additional file 1. Supplemental_Italian patent.