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DOE OSTI2021

Analysis of liquid petroleum using a laser-induced breakdown spectroscopy instrument

Here, a prototype analyzer for the direct LIBS analysis of nebulized liquid samples was developed and tested, particularly for the analysis of petroleum, organic solvents and aqueous solutions. The LIBS analyzer requires about 750 µl of liquid sample, 1 liter of N 2 gas, and 10 s of the analysis time to record 100 spectra. The limits of detection in oil and solvent are as low as 0.01–0.04 ppm for Li, Mg, and Cu. They increase for the difficult elements, such as Pb and Hg (7-10 ppm), Cl (250 ppm), and S (~0.7%). The relative standard deviation of measuring 100 ppm vanadium in oil and solvent was 1.5%. The LIBS detection limits and repeatability are better than required by the standard method ASTM D5185 for the analysis of lubricating oils in ICP-OES. Several petroleum samples were analyzed by LIBS and the quantitative results for V, Ni, and Fe compared to the ICP-OES data. Light crude oils can be nebulized and analyzed directly. Medium crude oils require minimal dilution at least 1:1, otherwise errors of determination become large. Presumably, utilization of the internal standard and chemometrics can be useful to correct for the matrix effects. In addition to the trace element analysis, the LIBS prototype demonstrated ability to measure the hydrogen-to-carbon ratio in organic liquid samples. Several unidentified features were observed in the carbon spectrum. Their possible origin is discussed.

02 PETROLEUM↗

DOE OSTI2022

Non-destructive analysis of swelling in the EMPIrE fuel test

The European Mini-Plate Irradiation Experiment (EMPIrE) was designed to support the development and testing of a coated uranium-molybdenum (U-Mo) dispersion fuel for the conversion of select high-performance research reactors (HPRRs) to utilize low-enriched uranium (LEU). To aid in the development of the coated fuel form, the EMPIrE test included several plate designs and irradiated them in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) at a high meat power density (~21 kW/cm 3 ) and to high fuel particle fission densities (~6.4 × 10 21 fissions/cm 3 ). These conditions mimic the bounding conditions of the BR-2 reactor in Belgium, where a concurrent irradiation experiment was performed, and exceed those previously explored in dispersion U-Mo fuel plates. A local fuel swelling analysis, as determined through high-fidelity, post-irradiation mini-plate profilometry, was used along with statistical methods to non-destructively evaluate the overall performance and separate the effects of convoluted fabrication variables. While some effects observed with this non-destructive analysis were subtle, others had more significant, and possibly competing, effects on the fuel swelling behavior. In closing, these observations will be examined further with destructive examinations to more fully assess them as the fuel design is developed and qualified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Support of the Development of EPP Plus SMT Design Method (FY2020 Test Results)

Experiments in support of the development of the integrated Elastic–Perfectly Plastic (EPP) analysis and Simplified Model Test (SMT) design methodology continued in FY 2020. The previously developed single-bar SMT (SBSMT) technique was successfully extended to tubular shaped Alloy 617 specimens for evaluation of the effect of primary pressure load on SMT creep-fatigue (CF) properties at 950°C. A series of pressurized SMT tests were performed with various elastic follow-up factors and strain ranges. The results show that the primary pressure load reduces the SMT CF cycles. The reduction of SMT CF life due to primary load was found to be dependent on strain ranges, elastic follow up and test temperatures. These results are crucial for verification of the EPP strain range analysis in the EPP+SMT evaluation procedure. Additionally, hold time effect on the SMT CF life cycles is a determining factor for generating the EPP+SMT CF design curves. Lack of test data in the high cycle and low strain range region results in major uncertainty in developing the design curves. In FY2020, experiments were designed, and testing started on Alloy 617 at 950°C to fill the gap in experimental data. The on-going CF and additional SMT CF with elastic follow-up at low strain ranges will generate the information needed to verify the methods for extrapolation to long hold time and finalize the EPP+SMT design method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Advancements Toward ASME Nuclear Code Case for Compact Heat Exchangers

Our research team proposes to advance the state of the ASME section III code (nuclear service) for Compact Heat Exchangers (CHX). This work will improve the technical state of CHXs and lay the foundation necessary for these heat exchangers to be certified for use in nuclear service. During the course of this work, we will advance the understanding of the performance, integrity, and lifetime of the CHXs for use in any industrial application, making their use more attractive and accessible to the industry. We will do this by developing qualification and inspection procedures that utilize Non-destructive evaluation (NDE) and advanced in-service inspection techniques, with insight from the industrial utility leader EPRI. We have enlisted colleagues at MPR Associates (MPR), an elite nuclear code consulting firm, who are experts on the ASME section III code and who, with input from members of the ASME section III committee, will direct the testing and help develop a series of documents that define the rules and regulations for use of the CHX. Colleagues at North Carolina State University (NCSU) and Oregon State University (OSU) will conduct extensive tensile, creep, and fatigue experiments on diffusion bonded samples (manufactured by US-based Vacuum Process engineering) along with modeling using the elastic perfectly plastic assumptions and comprehensive full inelastic finite element analysis (FEA). This work will allow analysis by design and confidence in the strength of different internal structures. To ensure industry acceptance and long term confidence, team members at the University of Wisconsin–Madison (UW), University of Michigan (UM), Georgia Tech (GT), and the University of Idaho (UI) will extensively test prototypic heat exchangers manufactured by US-based manufactures CompRex and Vacuum Process Engineering (VPE), a leader in the development of advanced CHX. This testing will include the use of various working fluids (salt, sodium, helium, and sCO2) to evaluate operational issues as well as structural integrity under the most severe conditions. Post-test analysis of the tested CHXs coupled with pre/in-service/post NDE (ultrasonic and radiography) led by the Electric Power Research Institute (EPRI) will be incorporated into the development of the rules and regulations for their use in nuclear service.

42 ENGINEERING↗

DOE OSTI2022

Steady-State Irradiation of Characterized Instruments for Nuclear Thermal Rockets Using In-Pile Experiment Apparatus

The development of nuclear rocket technology is critical for further exploration of extra-terrestrial bodies. Nuclear rockets can provide various advantages over current chemical rockets. However, the harsh environment provides several challenges regarding instrument performance. Temperature extremes, radiation (reactor and space), and the inability to conduct traditional maintenance activities can lead to severe degradation in instrument performance. To better understand the effects of radiation on instrumentation, an irradiation campaign has been designed to test a set of instruments provided by a collaborator in the aerospace industry. Details concerning instrument performance before and after irradiation are provided to allow for an analysis of the tested instruments. Additionally, the design of the irradiation test is described in detail.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2021

Analysis and Integration of the Hydraulic Fracturing Test Site-2 (HFTS-2) Comprehensive Dataset

Hydraulic Fracturing Test Site-2 (HFTS-2) is a field-based research experiment performed in the Permian (Delaware) Basin. The unique aspect of this program was the acquisition of a unique, comprehensive, diagnostic dataset. Additionally, shorter parent wells drilled three years before the child wells offered clear distinction between the stages influenced by parent-child effects and the stages without any effects. The goal of this study was to analyze and integrate this comprehensive diagnostic dataset to understand the areal and vertical extent of hydraulic fractures (HF). The paper also provides insights on the effects of parent wells’ depletion on child well HF geometry based on various monitoring methods and subsurface models. Areal and vertical coverage for all HFTS-2 wells during stimulation and depletion was estimated based on analysis and interpretation of diagnostics and advanced modeling results. HFTS-2 diagnostics included microseismic (MS), pre- and post-stimulation logs and cores, bottomhole gauges, and fiber optic (FO) data. The diagnostics results (MS, FO, image logs) were integrated and used to calibrate subsurface models. Additional field tests were designed and implemented for depletion monitoring. The tailored program for monitoring depletion included vertical and slant well pressures, interference testing, and a vertical strain depletion trial. Areal Coverage: Conventional MS (and FO MS) were used to compute HF dimensions, which were compared with diagnostics (FO strain, gauge, image logs) observations and calibrated subsurface models. A post-production interference test did not show offset well communication. Vertical Coverage: Vertical coverage during stimulation was monitored using a vertical monitoring well. The stronger mechanical strain signals showed good correlation with MS event intensities, geomechanical properties, and gauge inferences. Vertical depletion was estimated based on vertical/slant well gauges and strain depletion tests. Parent-Child Effects: Diagnostics and calibrated subsurface models show asymmetry in child well fracture geometries for stages that overlap parent wells. Child well image logs serve as a good indicator for parent Downloaded from http://onepetro.org/URTECONF/proceedings-pdf/21URTC/2-21URTC/D021S031R004/2477423/urtec-2021-5241-ms.pdf/1 by Carol Worster on 28 February 2022 URTeC 5241 well HF tracking. Child well MS events had an eastward bias, in line with pre-stimulation image logs, and was confirmed by parent well frac hits. Novel/Additive Information: The dataset presents a unique, over-constrained problem space to compare independent techniques to arrive at HF metrics (i.e., stimulation height and/or half-length), unlike a single source dataset, in which calibration is done using available data to guide predictions. Here, the asymmetry in HF geometry seen in the stages influenced by parent-child effects offers unique insights into well spacing and landing, which are key capital decisions the unconventional resources industry is seeking to optimize.

58 GEOSCIENCES↗

DOE OSTI2023

Product Consistency Test Results for the LAW HPVR Glasses

This report summarizes the chemical analysis of Product Consistency Test (PCT) leachates received from Pacific Northwest National Laboratory (PNNL). The leachates are from a series of simulated nuclear waste glasses designated Low-Activity Waste High PCT and VHT Response (LAW HPVR) glasses that were designed and fabricated at PNNL. The glasses included both quenched and canister centerline cooled glasses. The reported data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. The elemental release for the study glasses is reported as normalized concentration (NC i ). NC i of several elements was computed for both the target and measured glass compositions, which were similar, resulting in no significant differences when computing the NC i values. The majority of the glasses exhibited NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment Plant (WTP) low-activity waste constraint of 4 g/L. Several blank solutions included with the study glasses had K, Na, and Si concentrations above the analytical detection limit. Additionally, several of the reference glasses that were included with the study glasses had Na concentrations that were higher than expected. The unexpected analyte concentrations in the reference glass and blank solutions were insignificant with respect to the reported NC i values, and are noted for completeness and their experimental relevance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Product Consistency Test Results for the LAW ML1 Glasses

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Product Consistency Test Results for the HLW HAlG Glasses

This report summarizes the chemical analysis of Product Consistency Test leachates received from Pacific Northwest National Laboratory (PNNL). The leachates are from a series of quenched simulated nuclear waste glasses designated High-Level Waste High Aluminum Glass that were designed and fabricated at PNNL. The reported data is provided to be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. The elemental release for the study glasses is reported as normalized concentration NCi. NCi of several elements was computed for both the target and measured glass compositions. Several of the glasses exhibited NC B , NC Li , and/or NC Na values that were greater than the Environmental Assessment (EA) benchmark values. Several of the leachates from approved reference material (ARM) glasses included with the study glasses had elemental concentrations greater than the expected ranges. Two water blanks contained measurable amounts of Si; measurements confirmed by rerun samples. One water blank also contained a measurable amount of Na.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Analysis of Air-Purifying Respirator (APR) and Powered Air-Purifying Respirator (PAPR) Cartridge Performance Testing on a Hanford AP Tank Farm Exhauster Slipstream: Volume 2 Raw Analytical Data

As the Tank Operations Contractor for U.S. Department of Energy operations at the Hanford site in Washington State, Washington River Protection Solutions (WRPS) is responsible for managing highly radioactive wastes stored in tanks at Hanford. WRPS tests air-purifying respirator (APR) and powered air-purifying respirator (PAPR) chemical cartridges commonly used at Hanford Tank Farms to determine the period of time that the cartridges would provide adequate performance for APRs and PAPRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from any vapors exiting headspaces in the storage tanks. The Occupational Safety and Health Administration (OSHA) Standard promulgated in Title 29 of the Code of the Federal Regulations (CFR) 1910.134(d)(3)(iii)(b)(2) specifies that for protection against gases and vapors, employers shall implement a schedule for cartridges to ensure that change-outs occur before the end of service life. The change schedule can be based on objective information or data that ensures cartridge change-outs occur before the end of their service life.[2-5] The primary function of the WRPS Cartridge Test Program is to obtain objective data to determine service lives for the APR and PAPR cartridges used at Hanford Tank Farms. WRPS contracted with Pacific Northwest National Laboratory to analyze the test data and offer an independent analysis and any recommendations. Volume 1 of this report summarizes data analysis of APR and PAPR cartridge testing on vapors from the AP tank farm exhauster. Previous testing of APR cartridges was conducted on the AP exhauster in June 2016. However, an AP exhauster upgrade was completed in September 2016. Two different APR cartridges from SCOTT Safety (Monroe, North Carolina) were assessed for the new AP exhauster source, along with two different PAPR cartridges—one from MSA Safety Inc. (Pittsburgh, Pennsylvania) and another from 3M (Maplewood, Minnesota). Volume 2 provides an introduction to the raw data, including analytical laboratory analysis results that supported the analysis and conclusions documented in Volume 1.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2021

Development of Digital Twin Predictive Model for PWR Components: Updates on Multi Times Series Temperature Prediction Using Recurrent Neural Network, DMW Fatigue Tests, System Level Thermal-Mechanical-Stress Analysis

The long-term operation (LTO) of nuclear power plant (NPP) beyond their original design life of 40 years, can lead to more material damage associated with cyclic fatigue under thermal-mechanical loading cycles and associated long-term exposure of reactor material to the deleterious reactor-coolant environments. However, under this LTO condition the reactor components can still safely operate but may require more frequent Nondestructive Evaluation (NDE) of reactor components. Frequent NDE requirement may lead to frequent shutdown of the NPP. This in turn can lead to power outage and additional NDE-inspection-cost related economic loss. The economic loss can be minimized by reducing uncertainty in life estimation of safety-critical pressure boundary components and by implementing more digital approach such as by using upcoming digital-twin (DT) technology for predicting the structural states (e.g., time and location dependent inside/outside thickness temperature, stress, strain, plastic deformation, etc.) and associated fatigue life of a component in real time. Towards this goal Argonne National Laboratory (ANL) with the sponsorship of DOE Light Water Reactor Sustainability (LWRS) program is working on the development of a DT framework that can be used for real time environmental fatigue prediction of reactor components. The DT framework is based on limited experiment-data, Artificial-intelligence (AI) – Machine-Learning (ML) - Deep-Learning (DL) based techniques and Multiphysics-computational-mechanics such as finite element (FE) based modeling tools. Towards this overall goal, following are some of the major contributions made during the FY21: 1) Multiple 82/182 dissimilar metal weld (DMW) specimens (both solid-weld and joint-weld representing the actual reactor multi-metal nozzles) were fatigue tested. The resulting fatigue lives were compared to the NUREG-6909 based best-fit and design fatigue curves. Additionally, the results of 52/152 DMW fatigue specimens (which were recently tested at Republic of Korea under the sponsorship of International Nuclear Energy Research Initiative - INERI program) were compared to the NUREG-6909 based best-fit and design fatigue curves. From the comparison of 82/182 and 52/152 DMW test data with NUREG-6909 best-fit curve, most of the reported test data fall way away from the NUREG-6909 suggested best-fit or mean curve. The NUREG-6909 suggested best-fit curve is the best-fit curve of austenitic stainless steel and due to lack of enough data on Nickel-based welds, this is currently being used for predicting the life of Nickel-alloy-based welded components. However, the above observation may require higher scaling factor (e.g., ASME suggested factor of 20 on cycles rather than the current NUREG-6909 suggested factor of 12 on cycles) for scaling the austenitic-stainless-steel best-fit-curve for estimating the design or safe-life of a welded component. Accordingly, for example, if a DMW component experience a strain amplitude of 0.6% the PWR-water life of the component would be 52 cycles instead of 85 cycles. However, more DMW tests are required to further ascertain the above-mentioned observations. 2) A system level CAD and finite element model were developed which consists of reactor pressure vessel (RPV), part of steam generator (SG), part of pressurizer (PRZ), hot leg (HL), and surge line (SL). This is with detailed nozzle geometry and thermal-mechanical material properties of different metals to simulate realistic thermal-mechanical stress under connected system global thermal-mechanical boundary conditions. 3) Different system level heat transfer analyses were performed with estimation of relevant heat transfer coefficients. The resulting data were used in subsequent system level thermal-mechanical stress analysis and for generating spatial-temporal training and validation data for a system level digital-twin based temperature predictor. Transient heat transfer analyses were performed considering thermal boundary condition under design-basis (DB) loading and EDF (Électricité de France) data-based grid-load-following (EDF-GLF) loading cycles. 4) System level thermal-mechanical stress analysis was performed for identifying damage-prone hotspots and for future extension of the model for cyclic state prediction. From the system-level model simulation under DB loading cycle it is found that HL and the SL nozzle that connect to the HL can experience significant stress and strain and could be one of the weakest links in the overall reactor coolant system (RCS). 5) An AI/ML based DT model was developed for multi-time-series temperature prediction at any inside/outside thickness locations of PWR pressure boundary components. This is by using Recurrent-neural-network (RNN) and keras machine learning libraries. The RNN model was validated against two laboratory test-based data sets with one obtained through ANL’s in-air fatigue test system and other through PWR-water test loop. The experimentally validated DT model further validated against FE model results to predict thermal scarification related spatialtemporal temperatures at random locations of a component. The well validated DT model was then used for demonstrating spatial-temporal temperature prediction under 100+ years of reactor operation subjected to combined DB, EDF-GLF and randomized grid-load-following (RANDOMGLF) loading Cycles. The expert-elicitation DT model framework was developed assuming field/input/process measurements can be available from a few existing plant sensors and can readily be used by the NPP operators. The above temperature prediction model will feed to the next-step stress analysis model based on which the life of a component can be predicted in realtime, which is one of our future works.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Hygrothermal Aging and Thermomechanical Characterization of As-Manufactured Tidal Turbine Blade Composites

This study investigates the hygrothermal aging behavior and thermomechanical properties of as-manufactured glass fiber-reinforced epoxy and thermoplastic composite tidal turbine blades. The blades were previously deployed in a marine environment and subsequently analyzed through a comprehensive suite of material characterization techniques, including hygrothermal aging, dynamic mechanical analysis (DMA), tensile testing and X-ray computed tomography (XCT). Hygrothermal aging experiments revealed that while thermoplastic composites exhibited lower overall water absorption (0.78% vs. 0.47%), they had significantly higher diffusion coefficients than epoxy (2.1 vs. 12.1 × 10 −13 m 2 s −1 ), suggesting faster saturation in operational environments. DMA results demonstrated that water ingress caused plasticization in epoxy matrices, reducing the glass transition temperature and increasing damping (112 °C to 104 °C), while thermoplastic composites showed more stable thermal behavior (87 °C glass transition temperature). Tensile testing revealed substantial reductions in ultimate strength (>40%) for both materials after prolonged water exposure, with minimal change in elastic modulus, highlighting the role of matrix degradation over fiber reinforcement. XCT image analysis showed that both composites were manufactured with high quality: no large voids or cracks were present, and the degree of misalignment was low. These findings inform future marine renewable energy composite designs by emphasizing the critical influence of moisture on long-term structural integrity and the need for optimized material systems in harsh marine environments. This work provides a rare real-world comparison of epoxy and recyclable thermoplastic tidal turbine blades, showing how laboratory aging tests and advanced imaging reveal the influence of material and manufacturing choices on long-term marine durability.

16 TIDAL AND WAVE POWER↗

DOE OSTI2020

Investigation of Process Emissions during Thermal Cutting of Clad Metals from the Nuclear Sector - 20183

Typically, reactor pressure vessels as well as steam generators comprising low alloyed steel with a cladding of stainless steel. Recently the permissible values for Cr (VI) in the air of the working areas have been reduced in Germany from 100 μg/m{sup 3} to 1 μg/m{sup 3} air. This results in the danger that even with low-alloy steel with claddings these limit values would be exceeded when using flame cutting technologies. In order to ensure compliance with the permissible limit values, exhaust gas systems with gas cleaning measures are provided. The process emission test shall replace the assumptions required for the design of these systems with measured values. NUKEM Technologies Engineering Services GmbH (NUKEM) has chosen autogenous flame cutting as an example of a thermal cutting system for process emission testing, as there are no proven values in the literature for the Cr (VI) and NiOx emissions produced. The autogenous flame cutting can be used for a variety of cutting tasks within decommissioning projects. In order to simulate the material to be cut, mock ups were generated and cut from information about the material specification for the reactor pressure vessel from existing nuclear power plants including possible claddings. The cutting length needed for analysis in the test will be maximum 100 mm. The airborne emissions from the thermal cutting process will be captured and analyzed as follows: - Dust emission: The generated particles will be collected by a hood equipped with strong suction device and suitable filter elements. The resulting filter is gravimetrically measured before and after the tests. - Particle Size Distribution: The airborne particle size distribution measurement will result in figures for the size and number of particles arising from the thermal cutting process. - Gas Analysis: The gas analysis enables the determination of e.g. CO{sub 2}, CO, O{sub 3}, NO, NO{sub 2}. - Cr (VI) and NiOx Analysis: For the Cr (VI) and NiOx analysis, samples of the filter residue will be taken. As a result Cr (VI) in relation to total Cr as well as the Ni content will be determined by ICP-optically emission spectroscopy and put in relation to the total dust amount. The process emissions test itself as well as the necessary gas analyses and spectroscopic analyses were carried out at the Institute of Materials Science (IW) of Leibniz Universitaet Hannover (LUH). The results of the process emission tests are used to confirm the design of the gas cleaning systems. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Geomagnetically Induced Current Field Test on Large Grid-Connected Power Transformers: Analysis, Model Development, and Simulations

Geomagnetic-induced current (GIC) flow in power grids can cause undesirable effects such as transformer overheating, harmonics, higher reactive power demand, etc. Many simulation models have been developed to study these effects, but real-world verification on modern transformer designs is rare. Here, this paper presents the first long-duration GIC field test in the U.S. performed on high-voltage, grid-connected transformers featuring winding clamps and tie rods instead of conventional tie bars. Field measurements were taken to evaluate GIC effects. These measurements also aided in developing and validating thermal and electromagnetic transient (EMT) models of the transformers. During the test, significant current and voltage distortions were observed along with considerable transformer reactive power losses. Analysis of the field measurements showed that the transformers’ hottest spot was at the inner windings, and their k-factors were close to factory test and software default values. Thermal simulations indicated that the transformers would not violate their thermal limits even for a GIC waveform that peaks at about 200 A/phase. EMT simulations revealed that increased transformer loading may reduce GIC-induced reactive power demand and harmonics in certain scenarios. The study also highlighted potential inaccuracies in using the k-factor method to calculate transformer reactive power losses.

EMTDC↗

DOE OSTI2024

Compressed baryon acoustic oscillation analysis is robust to modified-gravity models

Abstract We study the robustness of the baryon acoustic oscillation (BAO) analysis to the underlying cosmological model. We focus on testing the standard BAO analysis that relies on the use of a template. These templates are constructed assuming a fixed fiducial cosmological model and used to extract the location of the acoustic peaks. Such “compressed analysis” had been shown to be unbiased when applied to the ΛCDM model and some of its extensions. However, it has not been known whether this type of analysis introduces biases in a wider range of cosmological models where the template may not fully capture relevant features in the BAO signal. In this study, we apply the compressed analysis to noiseless mock power spectra that are based on Horndeski models, a broad class of modified-gravity theories specified with eight additional free parameters. We study the precision and accuracy of the BAO peak-location extraction assuming DESI, DESI II, and MegaMapper survey specifications. We find that the bias in the extracted peak locations is negligible; for example, it is less than 10% of the statistical error for even the proposed future MegaMapper survey. Our findings indicate that the compressed BAO analysis is remarkably robust to the underlying cosmological model.

Astronomy & Astrophysics↗

DOE OSTI2020

Accelerated Aging Humidity Chamber for Nuclear Grade HEPA Filter Media

A unique accelerated aging humidity chamber for simulating the natural aging process of nuclear grade high efficiency particulate air (HEPA) filter media is being designed, implemented, and tested. The nuclear industry currently lacks information regarding the shelf life and service life of HEPA filter media. The Department of Energy recommends disposing of HEPA filters 10 years after the manufacture date for filters operated under dry conditions. Studies have shown the tensile strength and water repellency of HEPA filter media diminish with age. The chamber's principal function is to sustain an elevated relative humidity condition for an extended duration specified by the Arrhenius aging model. To accomplish this, various design parameters were established: (1) consistently achieve uniform humidity conditions within the chamber to equally expose media sheets, (2) ensure the chamber is sufficiently sealed to mitigate the uncontrolled infiltration of air, (3) confidently secure the media sheets to avert the introduction of unnecessary stressing/creasing, (4) fabricate the chamber from clear polycarbonate material to allow inspection of the media sheets, (5) provide a safe and ergonomic design for personnel, (6) ensure the chamber design is replicable and relatively easy to fabricate/assemble. A humidity source, fan, PID controller, humidity and temperature transmitters, and data loggers are required for proper operation and control of the chamber. The humidity source must accommodate target values up to 95% relative humidity throughout the chamber. The three humidity and temperature transmitters are strategically spaced throughout the chamber to ensure uniform conditions. The 10 cubic feet per minute (CFM) fan selected intends to accomplish the desired air change rate inside the chamber of approximately 20 air changes per hour (ACH). The 2 plenums on the upper and lower portions of the chamber utilize a perforated design to uniformly distribute air, promote air mixing, and control the air velocity entering the chamber. Industrial blueprint hanging clamps are being retrofitted to be seated within the chamber drawers and firmly secure the nuclear grade filter media. The aged media will be evaluated using autopsy methodology, and the results obtained intend to help clarify the useful life of nuclear grade HEPA filters. Nuclear grade high efficiency particulate air (HEPA) filters are defined as disposable, extended-media, dry-type filters with a rigid casing enclosing the full depth of the pleats and have a minimum particle removal efficiency of 99.97%. [1] - HEPA filters are constantly exposed to stressors and subsequently have been shown to degrade as the filter ages. Properties such as tensile strength and water repellency, along with others, are negatively impacted. [2] - Due to the scarcity of naturally aged HEPA filters available, it is crucial to develop accelerated aging methods which effectively mimic the natural degradation effects. - After accelerated aging, the filters are subject to testing in the Axial Flow Large Scale Test Stand (ALSTS) and other analysis techniques such as the same qualification tests performed directly after being manufactured. - Quality data is necessary for future decisions regarding the lifetime of nuclear grade HEPA filters. - The objective is to develop a prototype accelerated aging humidity chamber for nuclear grade HEPA filter media. The chamber should expose the media to commonly occurring stressors to expedite the degradation process caused by aging. PID yields reliable and reproducible results: - CFD model correlates well with actual performance; - Further analysis is needed to complete characterization; - Future work for the chamber is still to be completed. The data shown in Figure 5 and Table 1 indicate the system is functioning as designed. As can be determined from the system response evaluation in Table 1, the PID controller is yielding reliable and reproducible results at each set point. As the set point increases, the time constant and 20% settling time increase accordingly. From the results in Table 1, it stands to reason that the CFD model correlates well with the actual performance of the system. Each system response parameter for the 90%RH* CFD model is slightly lower than the parameters collected from the physical model tested at an equivalent set point. This is to be expected considering the inlet in the CFD model is a constant supply of water vapor, whereas the physical model uses a PID controller to monitor the amount of water vapor needed and reacts accordingly. More analysis is still to be done on the prototype chamber. A pressure test will be conducted in order to determine the hourly leak rate of the chamber. This test will involve constant air flow into the chamber until the desired pressure is reached. A uniformity test will also be conducted in order to ensure each filter medium is being exposed to equivalent relative humidity levels. This test will involve an array of humidity meters strategically oriented inside the chamber to record any possible gradients. Future work for the chamber includes dehumidification abilities, tests while fully loaded with filters, and temperature control. Once completed, the autopsy team at Institute of Clean Energy Technology will conduct necessary accelerated aging studies as needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

A Spatially Explicit Individual-Based Modeling Approach to Evaluate the Cumulative Effects of Wind Energy Development on the Greater Sage-Grouse: Sensitivity Analysis and Validation of Key Parameters

To address the need for an approach to evaluate cumulative ecological impacts of wind energy development, particularly those on critical wildlife habitats and species, the authors developed a proof-of-concept, landscape-based, spatially explicit individual-based modeling (IBM) framework for populations of the greater sage-grouse (Centrocercus urophasianus) in Albany County, Wyoming, based on published and other available information on the life history of the species. This sage-grouse IBM estimates the spatial and temporal movements of sage-grouse based on modeled effects of habitat characteristics and developments on sage-grouse condition, survivorship, and reproduction. After completing the initial version of the model, we conducted a series of sensitivity analyses of key model parameters and validation of lek occurrences and spatial distributions to better understand performance of the model relative to observed values of important life history characteristics. The parameter sensitivity analysis was conducted as a series of one-factor-at-a-time tests in which the values of selected parameters were individually varied to evaluate the corresponding effects on model output. We examined the model’s response to changes in parameter values to gain insight into the overall robustness of the model, identify key driver parameters, and initiate qualitative validation. Five functions selected for sensitivity analyses were (1) the rate of change in body condition in relationship to habitat suitability index (HSI) of occupied habitat (condition change rate), (2) the effect of competition between individuals with overlapping home ranges expressed as a percentage reduction in HSI (competition factor), (3) value of condition at which survivorship probability is equal to zero (zero survivorship condition), (4) the relationship between female body condition and clutch size, and (5) the relationship between female body condition and nest success. We varied the parameter values to be tested for the sensitivity analysis around the original value used in the model (i.e., tested values were higher and lower than the model’s original value) to examine the effect of this variation on the predicted output of the model. Model outputs evaluated in the sensitivity analysis included (1) predicted age-class distribution (as measured by the percentage of immature individuals within the population); (2) predicted sex ratio; (3) predicted age-sex class distribution (percent of population in each of six age-sex classes, i.e., male and female juveniles, yearlings, and adults); (4) lifespan; (5) population size; and (6) spatial distribution. Of the five parameters analyzed, condition change rate, survivorship, clutch size, and nest success did not substantially affect any of the evaluated model outputs. For the predicted age-class distribution, the percentage of immature individuals predicted by the model for these four parameters ranged from 61.9% to 65.0%, which was reasonable with respect to the reference values of 51.4 to 57.8% in published studies. The ratio of females to males ranged from 1.09 to 1.14, which was comparable to the reference values of 1.2 to 3.0 presented in an existing study. The predicted mean lifespan ranged from 1.62 to 1.74 years, which appears to be reasonable with respect to the reference values of 0.9 to 1.1 years for sharp-tail grouse and greater prairie chickens presented by published studies. The predicted population size ranged from 1,877 to 3,140. This prediction would be reasonable with respect to our estimate of 5,000 yearlings and adults, which was based on the USFWS estimate for the state of Wyoming scaled to the number of leks in the county. The predicted spatial distribution was not substantially affected by the parameter values examined. In our sensitivity analyses, the only tested parameter that had a noticeable effect on model output was competition factor. Over the range of values tested for this parameter, predicted population size exhibited the largest range (from 1,833 to 4,737) of all parameters tested. As the competition factor increased, the predicted population size decreased. This reflects the effect competition had on the number of individuals that could co occur in high-quality habitat patches. Despite this apparently significant effect, the range of values appears to be reasonable with respect to our estimate of 5,000 yearlings and adults, which was based on the USFWS estimate for the state of Wyoming scaled to the number of leks in the county.

17 WIND ENERGY↗

DOE OSTI2024

Effects of Neutron Irradiation on the Bond Strength of Steel Embedded in Concrete

The structural integrity of the concrete biological shield (CBS) in light water reactors (LWRs) exposed to long-term radiation raises the question of the effects of fast-neutron and gamma irradiation on the bond properties of embedded carbon-steel reinforcement in concrete, which is critical to the transfer of in-service and accidental loads. The design and execution of a first-of-a-kind irradiation experiment conducted in the LVR-15 test reactor (Czech Republic) are presented along with preliminary post-irradiation examination and testing results. Preliminary analysis shows that the fast neutron flux affects the radiation-induced volumetric of concrete aggregate, that the gamma irradiation induces shrinkage of the chemically bonded water in cementitious hydrates and that the bond strength of steel in concrete at the tested irradiation conditions decreases moderately.

Le Pape, Yann↗