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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Review of Destructive Assay Methods for Nuclear Materials Characterization from the Three Mile Island (TMI) Fuel Debris

This report provides a summary of the literature review that was performed and based on previous work performed at the Idaho National Laboratory studying the Three Mile Island 2 (TMI-2) nuclear reactor accident, specifically the melted fuel debris. The purpose of the literature review was to document prior published work that supports the feasibility of the analytical techniques that were developed to provide quantitative results of the make-up of the fuel and reactor component debris located inside and outside the containment. The quantitative analysis provides a technique to perform nuclear fuel accountancy measurements

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

University of Illinois Temperature Sensors

This document summarizes background information and presents results related to temperature measurements in the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) University of Illinois Project 29609 irradiation. The objective of this test was to assess the radiation performance of ferritic alloys for advanced reactor applications. The FeCr-based alloy system is considered the lead alloy system for a variety of advanced reactor components and applications. Irradiations of FeCr alloy samples were performed using the Hydraulic Shuttle Irradiation System (HSIS) in the B-7 position and in a static capsule in the A-11 position of the ATR.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Site Environmental Report for Calendar Year 2013: DOE Operations at The Boeing Company, Santa Susana Field Laboratory, Area IV

This Annual Site Environmental Report (ASER) for 2013 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of the Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2013 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling. Due to the suspension of D&D activities in Area IV, no effluents were released into the atmosphere during 2013. Therefore, the potential radiation dose to the general public through airborne release was zero. Similarly, the radiation dose to an offsite member of the public (maximally exposed individual) due to direct radiation from SSFL is indistinguishable from background. All radioactive wastes are processed for disposal at DOE disposal sites and/or other licensed sites approved by DOE for radioactive waste disposal. No liquid radioactive wastes were released into the environment in 2013.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Annual Site Environmental Report, Department of Energy Operations at the Energy Technology Engineering Center – Area IV, Santa Susana Field Laboratory

This Annual Site Environmental Report (ASER) for 2014 describes the environmental conditions related to work performed for the DOE at Area IV of the Santa Susana Field Laboratory (SSFL). The ETEC, a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

Frazee, Brad↗
DOE OSTI2013

Site Environmental Report For Calendar Year 2012. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

This Annual Site Environmental Report (ASER) for 2012 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2012 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

Liu, Ning↗
DOE OSTI2012

Site Environmental Report for Calendar Year 2011. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

This Annual Site Environmental Report (ASER) for 2011 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2011 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

Liu, Ning↗
DOE OSTI2011

Site Environmental Report for Calendar Year 2010. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

This Annual Site Environmental Report (ASER) for 2010 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2010 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

Liu, Ning↗
DOE OSTI2015

Fracture Capabilities in Grizzly with the extended Finite Element Method (X-FEM)

Efforts are underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). A capability was previously developed to calculate three-dimensional interaction- integrals to extract mixed-mode stress-intensity factors. This capability requires the use of a finite element mesh that conforms to the crack geometry. The eXtended Finite Element Method (X-FEM) provides a means to represent a crack geometry without explicitly fitting the finite element mesh to it. This is effected by enhancing the element kinematics to represent jump discontinuities at arbitrary locations inside of the element, as well as the incorporation of asymptotic near-tip fields to better capture crack singularities. In this work, use of only the discontinuous enrichment functions was examined to see how accurate stress intensity factors could still be calculated. This report documents the following work to enhance Grizzly’s engineering fracture capabilities by introducing arbitrary jump discontinuities for prescribed crack geometries; X-FEM Mesh Cutting in 3D: to enhance the kinematics of elements that are intersected by arbitrary crack geometries, a mesh cutting algorithm was implemented in Grizzly. The algorithm introduces new virtual nodes and creates partial elements, and then creates a new mesh connectivity; Interaction Integral Modifications: the existing code for evaluating the interaction integral in Grizzly was based on the assumption of a mesh that was fitted to the crack geometry. Modifications were made to allow for the possibility of a crack front that passes arbitrarily through the mesh; and Benchmarking for 3D Fracture: the new capabilities were benchmarked against mixed-mode three-dimensional fracture problems with known analytical solutions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Nanostructured Fe-Cr Alloys for Advanced Nuclear Energy Applications

We have completed research on the grain-size stabilization of model nanostructured Fe14Cr base alloys at high temperatures by the addition of non-equilibrium solutes. Fe14Cr base alloys are representative for nuclear reactor applications. The neutron flux in a nuclear reactor will generate He atoms that coalesce to form He bubbles. These can lead to premature failure of the reactor components, limiting their lifetime and increasing the cost and capacity for power generation. In order to mitigate such failures, Fe14Cr base alloys have been processed to contain very small nano-size oxide particles (less than 10 nm in size) that trap He atoms and reduce bubble formation. Theoretical and experimental results indicate that the grain boundaries can also be very effective traps for He atoms and bubble formation. An optimum grain size will be less than 100 nm, ie., nanocrystalline alloys must be used. Powder metallurgy methods based on high-energy ball milling can produce Fe-Cr base nanocrystalline alloys that are suitable for nuclear energy applications. The problem with nanocrystalline alloys is that excess grain-boundary energy will cause grains to grow at higher temperatures and their propensity for He trapping will be lost. The nano-size oxide particles in current generation nuclear alloys provide some grain size stabilization by reducing grain-boundary mobility (Zener pinning – a kinetic effect). However the current mitigation strategy minimizing bubble formation is based primarily on He trapping by nano-size oxide particles. An alternate approach to nanoscale grain size stabilization has been proposed. This is based on the addition of small amounts of atoms that are large compared to the base alloy. At higher temperatures these will diffuse to the grain boundaries and will produce an equilibrium state for the grain size at higher temperatures (thermodynamic stabilization – an equilibrium effect). This would be preferred compared to a kinetic effect, which is not based on an equilibrium state. The PI and coworkers have developed thermodynamic-based models that can be used to select appropriate solute additions to Fe14Cr base alloys to achieve a contribution to grain-size stabilization and He bubble mitigation by the thermodynamic effect. All such models require approximations and the proposed research was aimed at alloy selection, processing and detailed atomic-level microstructure evaluations to establish the efficacy of the thermodynamic effect. The outcome of this research shows that appropriate alloy additions can produce a contribution from the thermodynamic stabilization effect. Furthermore, due to the oxygen typically present in nominally high purity elemental powders used for powder metallurgy processing, the optimum results obtained appeared as a synergistic combination of nano-size oxide particle pinning kinetic effect and the grain-boundary segregation thermodynamic effect.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modal Frequency Identification of Vibrational Response Measurements: Simulation/Experiment/Validation

Many critical systems operating in the open seas or in the air employ a variety of subsystems ranging from machinery for propulsion to airfoils for flight or even reactor components as simple as pipes carrying much-need cooling for control purposes. In these subsystems,whether they be aboard a surface ship or submarine or AUV or in-flight aboard a rocket or aircraft—critical components must be monitored in real-time for safety and in some cases survival. Mechanical devices operating in noisy environments create a challenging signal processing problem to estimate their vibrational signatures in real-time. In order to estimate the performance of a particular type of device from noisy vibration data, it is necessary to identify signatures that make it unique. Resonant (modal) frequencies emitted offer a signature characterizing its operation.

42 ENGINEERING↗
DOE OSTI2019

Multichannel Processing of Vibrational Response Measurements: An Application of Constrained Subspace Identification

Dynamic structural systems occur in a wide variety of applications such as autonomous ve-hicles employed militarily for battlefield management or non-military tasks like tracking awildfire in the hills of a drought-ridden area. Many such systems operating in the air or atsea employ a variety of subsystems ranging from airfoils for flight or machinery for propulsionor reactor components as simple as pipes carrying cooling for control.

42 ENGINEERING↗
DOE OSTI2018

Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors

The overall summary of the waste management envelope discussed in this report is represented by the diagram shown in Figure S1. The streams that are included in this discussion include the following from reactor systems in which chlorine-based salts with and without processing and fluorine-based salts with and without processing are being considered: 1) Off-gas streams (e.g., particulates, aerosols, reactive gases, 3 H, water, residual halides, N 2 , O 2 , noble gases); 2) Salt waste streams (e.g., chloride- or fluoride-based); 3) Separated salt streams (e.g., 7 Li recycle, 37 Cl recycle, eutectic salts, fission products); 4) Metal waste streams (e.g., reactor components, plenums); 5) Carbon waste streams (e.g., graphite, SiC); 6) Decommissioning and decontaminating (D&D) waste streams; 7) Operating waste streams

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Infiltration Data Package for the E-Area Low-Level Waste Facility Performance Assessment

The infiltration data package contains the input parameters, the cap design and material properties assumptions, and the modeling results for the Hydrologic Evaluation of Landfill Performance (HELP) infiltration model simulations performed in support of the E-Area Low-Level Waste Facility Performance Assessment (PA). The infiltration estimates establish the upper boundary condition for the PORFLOW vadose-zone model and GoldSim model simulations for the following E-Area disposal unit types: Slit and Engineered Trenches, Low-Activity Waste Vault (LAWV), Intermediate-Level Vault (ILV), Componentin- Grout (CIG) special waste form trench segments, and the Naval Reactor Component Disposal Areas (NRCDAs). The infiltration data package builds upon relevant, foundational PA technical reports and memoranda from the past 15 years and is supported by three important components: the HELP model input parameter datasheets, HELP model input and output filenames and directory structure, and infiltration rates as a function of time for each scenario for each disposal unit type. For Slit and Engineered Trenches, a single best-estimate Bahia-grass case of two-percent slope and 585- foot slope length represents an upper bound on intact closure-cap infiltration rates over a 10,000-year period. Building upon the single, bounding, and intact case, two different approaches were considered for incorporating the effect of localized cap subsidence on infiltration rates in the PORFLOW flow and transport model simulations. In the first approach, a Monte Carlo probabilistic model was developed to calculate infiltration rates for 0.54%, 2%, 3.6%, and 4.9% subsidence scenariosa reflecting historical and future non-crushable content for the Slit and Engineered Trenches. The resulting blended (spatially and slope-length averaged) infiltration rates become an inflow upper boundary condition for the PORFLOW vadose zone model simulations. In the second and preferred approach, a Monte Carlo rectangle packing algorithm (Danielson, 2019) and a simplified equation for the total mass flux of water draining into a subsided compartment (Equation 1) will be employed to implement a weighted blending of radionuclide fluxes to the water table using the results of deterministic PORFLOW vadose zone simulations representative of specific subsidence cases. Finally, HELP-model results from sensitivity studies of the intact infiltration case were fit to a log-logistic function to generate infiltration profiles over a 10,000-year period for most-optimistic, more-optimistic, best-estimate, more-pessimistic, and most-pessimistic cases, for both the intact and four subsidence scenarios. For the LAWV and IL vaults, infiltration rate profiles for a 10,000-year period were developed for both an on-vault (above the concrete/metal vault roof) and an off-vault (10-foot soil zone adjacent to vault walls) scenario. The purpose of the off-vault simulations was to confirm that subsurface runoff from the concrete vault roof will adequately drain through the lowermost backfill layers adjacent to the vault walls. The LAWV and ILV infiltration profiles assume collapse of the concrete roof at relative Years 2,905 and 7,100, respectively. The identical Bahia-grass case with two-percent slope and 585-foot slope length assumed for the Slit and Engineered Trenches will represent an upper bound on intact infiltration rates for the NRCDAs and CIG special waste form as well. While the NRCDA casks are assumed to remain structurally intact (i.e., non-assumed to collapse 300 years after the end of operations. Because both rows of the existing CIG trench segments in ST23 (formerly CIG01) extend across the planned cap crest, the subsided CIG trench segments are treated conservatively in the HELP model as a 100%-subsidence case with zero subsurface run-on. The CIG special waste form trench segment infiltration profile, therefore, consists of an intact-cap period from relative Year 100 to Year 300, followed by a 100%-subsidence period from relative Year 300 to Year 10,100.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Initial study of notch sensitivity of Grade 91 using mechanisms motivated crystal plasticity finite element method

This report describes the development of a crystal plasticity finite element method (CPFEM) model for the multiaxial creep rupture of Grade 91 steel. The report demonstrates that core blocks in heat pipe reactor designs will see significantly different stress states than those typically found in conventional advanced reactor components where structures can often be represented as thin-walled pressure vessels. Current design practice uses effective stress measures calibrated to biaxial creep test data to account for the effect of stress multiaxiality on rupture life. Because these models rely on biaxial data they may not be accurate for the truly 3D, triaxial states found in heat pipe microreactor core blocks. A physically-based model, like the CPFEM model developed here, may be able to more accurately represent creep rupture under triaxial stresses when compared to conventional, empirical methods. This report describes the development, implementation, and validation of such a model and then applies the model to predict the effect of stress triaxiality on rupture in Grade 91 steel. The model results are used to assess the effective stress measures currently used in high temperature design codes and makes specific recommendations on improving these effective stresses to better account for the effect of 3D stress states on rupture in Grade 91. The validation and then implementation of these recommendations in design codes and standards could lead to safer, more effective heat pipe microreactor core block designs.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components. Third Annual Progress Report

Additive manufacturing (AM) of high-strength corrosion resistance alloys for nuclear energy applications, such as stainless steel and Inconel, is currently based on laser powder bed fusion (LPBF) process. Some of the challenges with using LPBF method for nuclear manufacturing include the possibility of introducing pores into metallic structures. Probability of crack initiation at the pore depends on size, shape, and orientation of the defect. Pulsed Infrared Thermography Imaging (PIT) provides a capability for non-destructive evaluation (NDE) of sub-surface defects in arbitrary size structures. The PIT method is based on recording material surface temperature transients with infrared (IR) camera following thermal pulse delivered on material surface with flash light. The PIT method has advantages for NDE of actual AM structures because the method involves one-sided non-contact measurements and fast processing of large sample areas captured in one image. Following initial qualification of an AM component for deployment in a nuclear reactor, a PIT system can also be used for in-service nondestructive evaluation (NDE) applications. In this report, we describe recent progress in enhancing PIT capabilities in detecting microscopic subsurface defects in metals, and classifying shapes and orientation of pores in thermal images. For detection of microscopic defects in PIT imaging data, we have developed Spatial Temporal Denoised Thermal Source Separation (STDTSS) unsupervised machine learning (ML) image processing algorithm. We show that flat bottom hole (FBH) defects as small as 200µm in SS316 and IN718 specimens, can be detected with STDTSS algorithm. To the best of our knowledge, these are the smallest detected defects which are reported in literature. For classification of defects shapes, we have previously developed thermal tomography (TT) algorithm to obtain depth reconstructions of material defects from data cube of sequentially recorded surface temperatures. However, interpretation of TT images is non-trivial because of blurring with increasing depth. To address this challenge, we have developed a deep learning convolutional neural network (CNN) to classify size and orientation subsurface defects in simulated TT images.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

ASME Code Revisions to Incorporate 316H and Alloy 617 Viscoplastic Constitutive Models to Section III, Division 5 and Code Case N-898

This report provides a final status update on work to develop and implement two new constitutive models for 316H stainless steel and the Ni-based Alloy 617 in Nonmandatory Appendix Z to the Section III, Division 5, Subsection HB, Subpart B ASME Boiler & Pressure Vessel Code rules covering the design and construction of Class A high temperature nuclear reactor components. This report summarizes the objections of the overall project and provides the final versions of the constitutive models proposed for incorporation into the ASME Code. The report also provides an update on the balloting status at ASME of the two proposed constitutive models. As of the time of writing (July 2022) the models are on-track to be approved by ASME after the August 2022 Code Week. If so, this will mean the 316H model will be published in the 2023 edition of the Code and the A617 model available as part of a revised Code Case immediately

36 MATERIALS SCIENCE↗
DOE OSTI2024

Near Net Shape Workshop Report

In 2021, President Biden issued Executive Order (EO) 14017, spurring the federal government to build more secure and diverse U.S. supply chains, including energy supply chains, to ensure economic prosperity and national security (White House 2021). In response to EO 14017, the U.S. Department of Energy (DOE) published America’s Strategy to Secure the Supply Chain for a Robust Clean Energy Transition (Igogo 2022) to outline the challenges and opportunities facing the energy supply chain along with key strategies to secure America’s position as a clean energy superpower in the years and decades to come. The expansion of domestic manufacturing capabilities is identified as one of the common risks and vulnerabilities for which DOE should develop comprehensive strategies to secure supply chains for the clean energy transition. The report states that “examples of manufacturing capabilities of concern include … large iron and steel castings for wind turbines, hydropower turbines, and nuclear reactor components.”

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗