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DOE OSTI2024

Summary of LWRS Research in Addressing RPV Research Gaps in NRC EMDA Report

Reactor Pressure Vessels (RPVs) are critical components in nuclear reactors, housing the reactor core and coolant under extreme conditions of temperature, pressure, and radiation. These harsh environments contribute to the degradation of RPV materials over time, presenting challenges for extending reactor operations beyond their original design lifespans. The NRC Expanded Materials Degradation Assessment (EMDA) report volume 3 have been instrumental in guiding research to support extending the operational life of light water reactors (LWRs) up to 80 years. It provides a comprehensive framework to address technical challenges related to aging and degradation mechanisms in RPVs. The LWRS program has played a key role in advancing this research, supporting projects such as the UCSB ATR-2 Experiment, material testing from Zion and Palisades reactors, and the development of advanced mini-compact tension testing techniques. These efforts have been crucial in identifying and addressing gaps in our understanding of RPV aging, contributing to the successful subsequent license renewals of eight LWR units in the U.S. The EMDA report volume 3, built on the Phenomena Identification and Ranking Table (PIRT) analysis from earlier versions of the EPRI Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs), provides a detailed assessment of RPV degradation mechanisms. However, as EPRI has updated the MDM and IMT, it is important to revisit research priorities and methodologies to reflect these changes. The revised MDM and IMT may introduce new factors affecting long-term RPV performance and safety, highlighting the need for continued research and updated guidance to ensure the reliable and safe operation of reactors beyond 80 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Metallurgical Analysis of the High Flux Isotope Reactor (HFIR) Carrier Lifting Bails (Rev.1)

The dissolution rates of the aluminum alloys in the High Flux Isotope Reactor (HFIR) element carriers and the Material Test Reactor (MTR) L-bundles in the H-Canyon facility have been identified as the possible cause of extended dissolutions that result in significant time and financial expenditures. A study, carried out by Savannah River National Laboratory (SRNL) to determine relationships between the dissolution rates and the metallurgical properties of the aluminum alloy materials of construction of the HFIR carriers and the L-bundles, considered the dissolution rates of aluminum alloy (AA) series 1100, 6061, and 6063. The study determined that the aluminum alloy compositions played a principal role in the dissolution rate of the carrier/bundle components. Higher dissolution rates were correlated with lower concentrations of the minor element additions in the alloys and with specific element concentrations. Aluminum alloys 1100 and 6063 were found to have similar dissolution rates that were approximately two orders of magnitude (100X) greater than those of AA6061. Based on the results of the dissolution behavior study, a Technical Assistance Request (TAR) was first issued to determine if the replacement of AA6061-T6 with AA6063-T6 is feasible for the HFIR carrier lifting bails. A Technical Task Request was then issued to consider AA6063-T5 as well as other alloys to improve possible supply chain issues. The metallurgical properties of the L-bundle (specifically the end caps) were not evaluated in this report because L-Bundle drawings already allow for the use of AA6063-T6 in all structural components. The HFIR carriers are composed of thin-walled components with significant surface areas that allow for relatively quick overall dissolution times. Conversely, the carrier lifting bails and the supporting constituents are composed of solid bars and thick plate regions with relatively small surface areas that experience longer overall dissolution times. While the MTR L-bundle design includes allowances for the materials of construction to be either AA6061-T6 or AA6063-T6, the HFIR carriers are specified to be constructed fully with AA6061-T6 alloy. This report analyzes the recommendations of the dissolution behavior study to replace the materials of construction of the HFIR carrier lifting bails. The analysis considers the operational requirements of the lifting bail and its supporting structures. To decrease dissolution times, the analysis considers direct replacement of the material as well as reductions in the thicknesses of the components to decrease the mass of the elements. Material reductions are considered on options for using either AA6061 and/or AA6063. The calculations are based on specifications from the American Society of Mechanical Engineer (ASME) and The Aluminum Association, Inc. design codes. The analysis finds that direct replacement of the lifting bail material of construction with AA6063-T6, and AA6063-T5 as well as reductions in the dimensions of the lifting bail components are acceptable. Note that this study considers the structural suitability of the alloys. It does not consider their dissolution rates in the dissolvers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Hyperspectral imaging for real-time waste materials characterization and recovery using endmember extraction and abundance detection

Hyperspectral imaging, combined with advanced spectral unmixing techniques and artificial intelligence, offers a powerful solution for improving material identification and classification. Here, this study evaluates the effectiveness of the pixel purity index and the sequential maximum angle convex cone algorithms in extracting and validating spectral signatures from pure samples of paper components (cellulose and lignin) and plastic (polypropylene). Principal-component analysis showed that both algorithms captured nearly all relevant variance for the tested materials. Spectral signatures were compared using the spectral angle mapper, revealing high similarity in the short-wave infrared region and greater variability in the visible near-infrared range. The methodology was then applied to a disposable coffee cup to detect and quantify mixed materials, accurately estimating material abundance and object area with less than 1% error. This approach enhances material classification, supporting product verification, quality control, and automated sorting for sustainable waste management and resource recovery.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Product Consistency Test Results for the LAW Phase 5 Glasses

This report provides the results of the Vapor Hydration Test solutions and the Product Consistency Test leachates from the Low-Activity Waste Phase 5 glasses, a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory. The series included quenched and canister-centerline cooled versions of the glasses. These data will be used in the development of enhanced property/composition models for waste glass vitrification at Hanford. Several of the blanks had detectable amounts of sodium and/or silicon. The measured concentrations (mg/L) of boron, sodium, and silicon for the low-activity test reference material glass included with the Product Consistency Tests were much lower than the expected low-activity test reference material test results. A review of the Product Consistency Test data indicated that there was generally little difference between the normalized values based on targeted or measured glass composition. Heat treatment had various impacts on the normalized concentration values. Most of the glasses had NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment and Immobilization Plant low-activity waste constraint of 4 g/L. The release rates for boron, sodium, and silicon were highly correlated for the study glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Temperature-Time Modeling of Spent Fuel Cladding in Dry Storage Casks

This report uses thermal and decay heat modeling to investigate spent fuel performance and the potential for cladding to anneal in dry storage conditions. Annealing is an important feature to investigate in SNF cladding because it has a direct relationship to cladding response in storage, transportation and disposal conditions. Annealed cladding may have lower strength than unannealed cladding, however its increased ductility would provide better protection against rupture in high strain rate situations such as severe accidents. These consequences are the driver for this work however are outside of the scope of this report. The modeling focused on three representative storage systems, the TN-32B, MAGNASTOR with TSC-37 canister, and NUHOMS AHSM with a 32PTH2 canister. This covers the vertical dual-purpose, vertical ventilated and horizontal ventilated casks respectively. Decay heat modeling using high and low enrichment assemblies was used to bound the decay heat curves that might be expected in dry storage. To bound the temperature relationship, the storage casks were modeled starting at the design basis heat loads with heat decaying through time. Although the results are bounding there is not an attempt to maximize conservatism, rather the intent to form a reasonable upper limit on temperature that will be broadly applicable to the U.S. cask fleet. This will allow materials testing to focus on relevant conditions for annealing that may affect the U.S. spent fuel inventory. The results show a clear dependence on heat load pattern in time in Figure S-1 and Figure S-2. This dependence is due to the different assembly decay heat curves for different assemblies in preferentially loaded casks. It shows the need for careful decay heat modeling when examining in service fuel temperatures that are less than the cask design basis heat load. The body of the report shows percent cladding cutoffs of 300°C and 350 °C as well. These results can be used to inform testing and conclusions about cladding performance through time and the potential for cladding annealing during dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multi-site Milling Strategy Reveals Significant Variation in Biomass Composition of Switchgrass ( Panicum virgatum ) Grown at Ten Locations

Cell wall composition influences biomass use as a forage and as a feedstock for biofuel and chemical conversion. To examine the influence of environment on composition of switchgrass (Panicum virgatum L.), we utilized a multi-environment experiment consisting of clones of switchgrass genotypes grown at up to ten locations in the continental US. We tested the influence of different milling treatments on biomass composition trait predictions via near-infrared reflectance spectroscopy (NIRS). We found that most compositional trait predictions (29/34) were significantly different (P < 0.05) when a single lot of biomass was subjected to disparate milling treatments, i.e., knife milling vs. knife milling with an additional cyclone milling. Further, depending on the plant material tested, three to eight compositional trait predictions vary (P < 0.05) when identical biomass was knife milled at different sites followed by cyclone milling at a single site, including for traits such as Klason lignin, nitrogen, and carbon. In some cases, variation due to milling site exceeded environmentally induced compositional variation of a single switchgrass genotype grown at different sites. From these observations, we recommend a protocol with two sequential millings that decouples growth environment from a particular mill. Utilizing this approach, we found that 46/46 biomass composition traits from the warm season herbaceous forage and switchgrass bioethanol NIRS equations vary significantly (P < 0.001) in clones of a switchgrass genotype (WBC) grown at ten sites, with the growth site representing the largest average source of variation (41%). This multi-site milling approach can be used to examine environmental and gene-by-environment influences on composition with the goal of optimizing cell wall composition in different environments for biomass utilization.

09 BIOMASS FUELS↗
DOE OSTI2023

Binder Jet Additive Manufacturing Process and Material Characterization for High Temperature Heat Exchangers Used in Concentrated Solar Power Applications

The U.S. Department of Energy’s (DOE) Sunshot 2030 initiative has a goal of reducing the cost of concentrating solar power (CSP) to 5 cents per kWh for baseload power plants. One of the potential pathways to this goal includes a reduction in the cost of the supercritical CO2 (sCO2) power block to 0.9 cents per kWh. Recuperators—high and low temperatures, used in the sCO2 power cycle, contribute to >50% of the cost of the power cycle. This work studies the feasibility towards a ≥10% cost reduction for High Temperature Recuperators (HTR) used in the sCO2 power cycle. One way to address the cost reduction is by leveraging low-cost additive manufacturing, specifically, Binder Jet Additive Manufacturing (BJAM) to 3D print HTRs at scale. This study focuses on the development of a BJAM process towards 3D printing HTR cores using Stainless Steel alloy 316L (SS316L). To evaluate the suitability of the BJ process towards the HTR, high level specifications of the application are translated to materials capability requirements. Subsequently, at-temperature materials testing is conducted on as-printed and sintered additively manufactured coupons. Data from the coupons are compared against cast and wrought SS316L data obtained from the literature. Results show that the tensile properties from the BJ process compare well against cast properties. Furthermore, a baseline analysis of creep testing data is established for the BJ process, and insights are drawn from the results towards future improvements of the process.

14 SOLAR ENERGY↗
DOE OSTI2022

An Improved Pathway to Ethylene Glycol via CO 2

This report describes our efforts to develop a partial electrochemical oxidation process to produce formaldehyde selectively from methanol. The effort examined numerous potential anode materials in pursuit of 70% Faradaic efficiency for conversion of methanol to formaldehyde. Among the various electrode materials tested herein both Pt and PtRu were identified as the most promising catalysts in this process as the Faradaic efficiency obtained in these cases are 40% and >70% respectively. Several other 3d transition metals, Pt group metals and their alloys were also studied in which the Faradaic efficiency was obtained within 5-30%. In case of Pt and PtRu bulk metal electrode shows higher selectivity to formaldehyde compared to their nanostructured materials. Based on the literature, nanostructured Pt and PtRu are considered as active catalysts for methanol fuel cells, thus, lower selectivity to formaldehyde with nano Pt and PtRu in our case was attributed to their preferential oxidation of methanol to CO 2 . Severe catalyst deactivation was noted when 0.5 (M) sulfuric acid was used as the electrolyte irrespective of catalyst formulation. On the other hand, much higher activity and selectivity for formaldehyde was noted when alkaline electrolyte was employed during oxidation. Considering selective electrochemical oxidation of methanol to formaldehyde is an area which is underexplored as of today, our findings about catalyst formulation and process condition is noteworthy in this direction. Both bulk metal electrodes of Pt and PtRu may be cost prohibitive and thus present challenges to the scale up this technology. Therefore, we recognize that additional efforts are needed to develop catalysts with high activities and Faradaic efficiencies similar to bulk electrodes for the electrooxidation of methanol to formaldehyde.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Mechanical Loads on Spent Nuclear Fuel in the General 30 cm Package Drop Scenario

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) research program is guided by the high-level goal of closing prioritized knowledge gaps related to spent nuclear fuel (SNF) storage and transportation, which are summarized by Saltzstein et al. (2020). One of the high-priority knowledge gaps is the identification and quantification of mechanical loads that are expected to affect SNF during normal conditions of transportation and storage to inform the range of physical SNF test programs. This report uses modeling and analysis methods to estimate the mechanical loads on spent nuclear fuel (SNF) in the general 30 cm package drop scenario. The drop scenario assumes impact limiters are in place in the transportation configuration and the impact surface is perfectly rigid. The goal of this analysis is to consider the universe of potential mechanical loading conditions that can happen to SNF and present the results in a manner that is useful for materials testing, decision making, and regulatory rule making purposes. This study uses validated finite element models and methods to perform a broad parametric study of key variables that can affect the mechanical loads on SNF during a hypothetical package free drop scenario. Physical drop test data from a cask and fuel assembly drop test campaign is the basis for model validation. Additionally, the results of the parametric study are used to inform a damage model, which uses multiple nonlinear regression to estimate the relationships between input variables and output response. The parametric finite element analyses consider thousands of input variable combinations, while the damage model estimates millions of combinations. The breadth of this study provides confidence that the potential range of mechanical loads that SNF might experience during the general 30 cm package drop scenario are characterized well enough to consider this knowledge gap closed. While this report documents the overall peak values calculated in this study, the 95 th percentile values, the histograms, and the observed trends are equally important. This study covered a large range of SNF temperatures, room temperature to 300°C, and burnups, 10 GWd/MTU to 62 GWd/MTU. Each temperature and burnup combination has a different cladding yield strain, so it is more meaningful to summarize the calculated cladding strain response as its factor of safety, which is defined relative to the yield strain. The factor of safety is calculated as the yield strain divided by the peak cladding strain. A factor of safety greater than unity indicates that the cladding remains below yield, whereas a value less than unity is indicative of plastic deformation. In all cases of this study a safety factor over 1.0 was calculated, although in the most limiting case at 300°C the safety factor was only 1.01, which suggests that yielding could occur when additional loads like rod internal pressure are included. When the temperature is restricted to 200°C the limiting safety factor increases to 1.28, which has significant margin to accommodate internal pressure and potential local cladding defects that could cause a local stress concentration. An important trend in the calculated fuel rod mechanical loads is that the 2 nd highest loaded fuel rod in an assembly tends to be significantly lower than the highest loaded rod. The implication is that even if one rod in an assembly experiences a failure the loads would have to be significantly higher to cause two or more rods to fail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Advanced Test Reactor Long-Term (20-Year) Operational Strategy

The Department of Energy Office of Nuclear Energy (DOE-NE) and the Naval Nuclear Propulsion Program (NNPP) have identified the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) as a critical need for continued thermal irradiation testing capability to support and advance U.S. commercial and naval nuclear power systems. The ATR is currently the only suitable test reactor operating in the world that can provide these capabilities at volume. ATR must, then, plan to operate for at least the next 20 years. As of the beginning of 2024, ATR has operated for 57 years, with many of the original reactor structures, systems, and components installed and operated for 60 years. Some ATR Complex and reactor support equipment that is currently in use was installed 75 years ago with the startup of the Materials Testing Reactor (MTR).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

X-Ray Microtomography Measurements of Coevolving Particle Size and Shape

This study measures simultaneous changes in particle size and shape during grain crushing and proposes quantitative metrics to track their coevolving statistical distribution. Idealized granular materials with widely different initial particle shapes (i.e., spheres, quasi-ellipsoids, plates, and rods) are compressed oedometrically at vertical stress levels sufficient to induce particle breakage. X-ray microtomography and digital image analysis are then used to track changes in particle-scale characteristics. The observations indicate that more irregular particles tend to exhibit a higher degree of breakage compared to more spherical grains subjected to an equivalent stress level. Despite these differences, the measurements also indicate that all the tested particle sets approach a common attractor (expressed in terms of aspect ratio, flatness, and elongation), regardless of their initial morphology. A new shape evolution index is proposed to quantify such trends. With this new metric, it is shown that in all the tested materials, the particle shape evolves faster than size, especially for the more regular morphologies, thus implying that the shape tends to reach its attractor point before the grading reaches its ultimate particle size distribution. Furthermore, this finding underpins an ultimate stage of crushing-driven compression that occurs with particle size reduction but minimal shape changes (i.e., self-replicating shapes due to fractures).

Oedometric compression↗
DOE OSTI2021

Thermal-Mechanical Elastic-Plastic and Ductile Failure Model Calibrations for 6061-T651 Aluminum Alloy from Plate

Numerical simulations of metallic structures undergoing rapid loading into the plastic range require material models that accurately represent the response. In general, the material response can be seen as having four interrelated parts: the baseline response under slow loading, the effect of strain rate, the conversion of plastic work into heat and the effect of temperature. In essence, the material behaves in a thermal-mechanical manner if the loading is fast enough so when heat is generated by plastic deformation it raises the temperature and therefore influences the mechanical response. In these cases, appropriate models that can capture the aspects listed above are necessary. The matters of interest here are the elastic-plastic response and ductile failure behavior of 6061-T651 aluminum alloy under the conditions described above. The work was accomplished by first designing and conducting a material test program to provide data for the calibration of a modular $J_2$ plasticity model with isotropic hardening as well as a ductile failure model. Both included modules that accounted for temperature and strain rate dependence. The models were coupled with an adiabatic heating module to calculate the temperature rise due to the conversion of plastic work to heat. The test program included uniaxial tension tests conducted at room temperature, 150 and 300 C and at strain rates between 10 –4 and 10 3 1/s as well as four geometries of notched tension specimens and two tests on specimens with shear-dominated deformations. The test data collected allowed the calibration of both the plasticity and the ductile failure models. Most test specimens were extracted from a single piece of plate to maintain consistency. Notched tension tests came from a possibly different plate, but from the same lot. When using the model in structural finite element calculations, element formulations and sizes different from those used to model the test specimens in the calibration are likely to be used. A brief investigation demonstrated that the failure model can be particularly sensitive to the element selection and provided an initial guide to compensate in a specific example.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Simulation Analysis of Geometry and Material Effects for Dropkinson Bar

The reported research is motivated by the need to address a key issue affecting the Dropkinson bar apparatus. This unresolved issue is the interference of the stress wave reflected from the bar-beam boundary with the measurement of the stress-strain response of a material tested in the apparatus. The purpose of the wave beam that is currently connected to the bar is to dissipate the stress wave, but the portion of the wave reflected from the bar-beam boundary is still significant. First, we focused on understanding which parameters affect the reflected wave's arrival time at a strain gauge. Specifically, we used finite-element numerical simulations with the Sierra/SM module to study the effects of various bar-beam connection fixities, alternative wave beam materials, and alternative geometries of the Dropkinson bar system based on a monolithic design. The conclusion of this study is that a partial reflection always occurs at the bar-beam boundary (or, for a monolithic design, at a point where the bar geometry changes). Therefore, given a fixed total length of the bar, it is impossible to increase the reflected wave's arrival time by any significant amount. After reaching this conclusion, we focused instead on trying to minimize the energy of the reflected stress wave circulating up and down through the bar over a relatively long period of time (10 ms). Once again, we used numerical simulations with the Sierra/SM module to investigate the effects of various bar-beam connection fixities, alternative wave beam materials, and parameters of an asymmetric monolithic design of the bar-and-beam system. This study demonstrated that various parameters can significantly affect the energy of the wave reflections, with the difference between best and worst configurations being about one order of magnitude in terms of energy. Based on the obtained results, we conclude with concrete takeaways for Dropkinson bar users and propose potential directions for future research and optimization.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Technical Evaluation of Accelerated Basin De-Inventory Material Addition to Sludge Batch 11

Savannah River Nuclear Solutions has a need to discard spent nuclear fuel (SNF), currently stored in L Basin, to the Defense Waste Processing Facility (DWPF) for vitrification. The Department of Energy (DOE) has approved the Accelerated Basin De-inventory (ABD) Program for discarding SNF via transfers from H-Canyon to the Savannah River Site (SRS) Liquid Waste (LW) system. The first ABD discards will occur during the preparation of Sludge Batch (SB) 11. An initial impact evaluation of the LW flowsheet was performed by the Savannah River National Laboratory (SRNL) prior to the approval of the ABD Program. This evaluation addressed the LW downstream facilities based on the current H-Canyon flowsheet sequence for the average ABD discard. The flowsheet evaluation only included aluminum-clad SNF, specifically Materials Test Reactor (MTR) fuel and High Flux Isotope Reactor (HFIR) fuel similar to the planned SB11 discard. Following this evaluation, the flowsheet has been slightly altered to address (i) new nuclear criticality safety controls for DWPF that credit a higher amount of gadolinium as a neutron poison for all of the enriched uranium contained in a SB and (ii) potential additions of the H-Canyon neutralized fuel stream prior to the Low Temperature Aluminum Dissolution (LTAD) process in Tank 51. The early introduction of ABD material into the SB assembly process is being investigated to provide flexibility regarding transfer opportunities for H-Canyon to Tank 51 for SB11 and future sludge batches.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Evaluation of the Dissolution Behavior of L-Bundle End Caps and HFIR Fuel Carriers

The H-Canyon facility is currently using the 6.1D dissolver for the dissolution of Material Test Reactor (MTR) fuel and the 6.4D dissolver for the dissolution of High Flux Isotope Reactor (HFIR) fuel using mercury-catalyzed HNO₃ dissolution flowsheets. The processing strategy for both dissolvers involves the dissolution of multiple charges of fuel per batch. After the designated heating cycle, the dissolvers are opened, and the charging wells are probed to determine if the MTR or HFIR fuel has dissolved. If undissolved fuel fragments are beyond a certain height, the dissolver must be closed and heated for an additional amount of time to dissolve the remaining material. In recent MTR fuel dissolutions, “high probes” (i.e., excessive undissolved material) were frequently observed, which resulted in extended dissolution times. The suspected cause of the high probes was the incomplete dissolution of the L-Bundle End Cap, rather than the fuel or fuel bundles. The End Cap is hypothesized to be binding in the insert well and not dropping into the acid as the L-Bundle and fuel dissolve. Once the End Cap is dislodged by the probe and drops into the acid, the dissolution rate of the End Cap appears to be significantly reduced compared to the dissolution rate of the fuel and other parts of the L-Bundle. A similar issue has also been observed with the lifting bail on the HFIR fuel carriers. During HFIR fuel dissolutions, the lifting bail on the outer carrier has resulted in high probes due to incomplete dissolution. In one case, a partially dissolved bail was caught in one of the insert well holes which prevented the probe from going to the bottom of the well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Structural Characterization of Deployed Thermoplastic and Thermoset Composite Tidal Turbine Blades

The National Renewable Energy Laboratory worked with Verdant Power to manufacture, and characterize, novel thermoplastic composite blades on their Gen5d 5 m diameter turbines at the Roosevelt Island Tidal Energy (RITE) site in the East River in N.Y. to demonstrate a low-cost manufacturing process for marine energy structures. Verdant had designed, manufactured, and deployed epoxy thermoset composite blades on three Gen5d turbines in October 2020. At a maintenance cycle in May 2021, a Gen5d turbine equipped with the NREL thermoplastic blades was deployed and retrieved in October 2021. Modal, static and fatigue structural characterization was performed on both blade types before and after the deployment. The static and modal test results showed that the thermoplastic blades were slightly stiffer than the epoxy blades both dry and after the deployment. The epoxy blades had about 8% increase in strain at the applied load after the deployment, whereas the thermoplastic blade strains were not changed significantly. Ultimately the thermoplastic blade failed during fatigue testing due to a crack which is thought to have initiated between the internal foam and the laminate at the location of the internal instrumentation. The Verdant Power Gen5d epoxy blades did not have this internal instrumentation and did not have this failure, performing adequately through the operational period. A lack of design information and fatigue data for both the thermoplastic laminates and the adhesive used means that further work is needed to fully understand his failure and the root cause analysis is ongoing. This paper provides details on the thermoplastic blade manufacturing, materials, test methodology and results.

17 WIND ENERGY↗
DOE OSTI2021

Recombination of Hydrogen in the Iodine Reactors

The H-Canyon facility is currently dissolving spent nuclear fuel, including Material Test Reactor (MTR) and High Flux Isotope Reactor (HFIR) fuel. Dissolution of aluminum spent nuclear fuel produces hydrogen and other NOx gases. A theory long held by H-Canyon Engineering and Facility Technical Advisors that the silver nitrate-coated berl saddles, present in the iodine reactor, used for off-gas treatment were acting as a catalytic hydrogen recombiner was captured during an external review of the Accelerated Basin Deinventory program. Results from a separate SRNL project using a Raman Spectrometer to monitor the offgas from the dissolution process indicated a lack of the expected hydrogen in the off-gas stream. To test this hypothesis, a laboratory scale iodine reactor was assembled and filled with silver nitrate-coated berl saddles. Testing with this laboratory scale reactor confirmed the recombination of hydrogen when a simulated dissolver off-gas was passed through the reactor containing silver nitrate-coated berl saddles at 173-188 °C. Control experiments performed with uncoated berl saddles resulted in no change to the hydrogen concentration after passing through the reactor. The residence time of the gas in the reactor was varied to determine if residence time had an impact on the amount of hydrogen recombination occurring. Results from these experiments indicated that at the shortest residence times tested (~5 seconds) recombination of the hydrogen still exceeded 90%; however, the percent recombination did increase at longer residence times, reaching 97.9% or greater for residence times over 10 seconds. Finally, testing performed with a simplified off-gas composition containing only hydrogen and air gave similar results, indicating that the presence of NO and N 2 O gases in the stream do not play a role in the recombination reaction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Options, Initial Design Requirements, Estimated Costs, Reactor Commitments, and Potential Uses of a Graphite Leadout Type Experiment Supporting Various Commercial HTR Vendors

Multiple commercial High Temperature Reactor (HTR) vendors and nuclear graphite suppliers would benefit by collaborating on a new irradiation capsule(s) that would include graphite grades not included within the AGC Experiment. This new irradiation capsule(s) would provide data to answer vendor graphite licensing issues. Rather than spending money (and especially) time in designing separate irradiation capsules for each designer, the capsule(s) would be used for multiple graphite and composite designs to maximize efficiency and promote multiple HTR designs. However, the primary motivation for assisting vendors with this new irradiation capsule(s) is lack of availability in the existing Material Test Reactors (MTRs). Cost reduction is a secondary goal. A common, collaborative, capsule design can be achieved for graphite and composites due to the similarity of different grades. Irradiation, disassembly, shipping, and PIE costs would be cost-shared by all users. It is anticipated that interest would extend across all DOE campaigns (micro-Rx, SMR, GCR, MSR, etc.) due to the similar requirements for all graphite grades.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗