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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Readout electronics for the CMS Phase II Endcap Calorimeter: system overview and prototyping experience

Abstract The frontend readout system for the silicon section of the CMS Phase II Endcap Calorimeter faces unique challenges due to the high channel count and associated bandwidth, limited physical space, as well as radiation tolerance requirements. I will give an overview of the frontend electronics design and will discuss the recent experience obtained from the first test system that integrates the HGCROC2 readout ASIC, lpGBT, and VTRX+ in a realistic manner, linking together prototypes of the hexaboard, engine, and wagon boards.

Instruments & Instrumentation↗
DOE OSTI2023

Specifications and pre-production of n + -in-p large-format strip sensors fabricated in 6-inch silicon wafers, ATLAS18, for the Inner Tracker of the ATLAS Detector for High-Luminosity Large Hadron Collider

Abstract The ATLAS experiment is constructing new all-silicon inner tracking system for HL-LHC. The strip detectors cover the radial extent of 40 to 100 cm. A new approach is adopted to use p-type silicon material, making the readout in n + -strips, so-called n + -in-p sensors. This allows for enhanced radiation tolerance against an order of magnitude higher particle fluence compared to the LHC. To cope with varying hit rates and occupancies as a function of radial distance, there are two barrel sensor types, the short strips (SS) for the inner 2 and the long strips (LS) for the outer 2 barrel cylinders, respectively. The barrel sensors exhibit a square, 9.8 × 9.8 cm 2 , geometry, the largest possible sensor area from a 6-inch wafer. The strips are laid out in parallel with a strip pitch of 75.5 μm and 4 or 2 rows of strip segments. The strips are AC-coupled and biased via polysilicon resistors. The endcap sensors employ a “stereo-annulus” geometry exhibiting a skewed-trapezoid shapes with circular edges. They are designed in 6 unique shapes, R0 to R5, corresponding to progressively increasing radial extents and which allows them to fit within the petal geometry and the 6-inch wafer maximally. The strips are in fan-out geometry with an in-built rotation angle, with a mean pitch of approximately 75 μm and 4 or 2 rows of strip segments. The eight sensor types are labeled as ATLAS18xx where xx stands for SS, LS, and R0 to R5. According to the mechanical and electrical specifications, CAD files for wafer processing were laid out, following the successful designs of prototype barrel and endcap sensors, together with a number of optimizations. A pre-production was carried out prior to the full production of the wafers. The quality of the sensors is reviewed and judged excellent through the test results carried out by vendor. These sensors are used for establishing acceptance procedures and to evaluate their performance in the ATLAS collaboration, and subsequently for pre-production of strip modules and stave and petal structures.

Instruments & Instrumentation↗
DOE OSTI2023

The New Small Wheel electronics

The increase in luminosity, and consequent higher backgrounds, of the LHC upgrades require improved rejection of fake tracks in the forward region of the ATLAS Muon Spectrometer. The New Small Wheel upgrade of the Muon Spectrometer aims to reduce the large background of fake triggers from track segments that don't originate from the interaction point. The New Small Wheel employs two detector technologies, the resistive strip Micromegas detectors and the “small” Thin Gap Chambers, with a total of 2.45 million electrodes to be sensed. The two technologies require the design of a complex electronics system given that it consists of two different detector technologies and is required to provide both precision readout and a fast trigger. It will operate in a high background radiation region up to about 20 kHz/cm 2 at the expected HL-LHC luminosity of $\mathcal{L}$= 7.5 × 10 34 cm -2 s -1 . The architecture of the system is strongly defined by the GBTx data aggregation ASIC, the newly-introduced FELIX data router and the software based data handler of the ATLAS detector. The electronics complex of this new detector was designed and developed in the last ten years and consists of multiple radiation tolerant Application Specific Integrated Circuits, multiple front-end boards, dense boards with FPGA's and purpose-built Trigger Processor boards within the ATCA standard. The New Small Wheel has been installed in 2021 and is undergoing integration within ATLAS for LHC Run 3. It should operate through the end of Run 4 (December 2032). In this manuscript, the overall design of the New Small Wheel electronics is presented.

47 OTHER INSTRUMENTATION↗
DOE OSTI2025

Testing the limits of ITkPixV2: the ATLAS inner tracker pixel detector readout chip

The ITkPixV2 chip is the final production readout chip for the ATLAS Phase 2 Inner Tracker (ITk) upgrade at the upcoming High-Luminosity LHC (HL-LHC). Due to the extraordinarily high peak luminosity at the HL-LHC of 5 × 1034 cm-1 s-1, ITkPixV2 must meet significant increases in nearly all design requirements compared to the current ATLAS Inner Detector (ID), including a 10× increase in trigger rate, a 7.5× increase in hit rate, a 3× increase in radiation tolerance, and a 12.5× decrease in pixel current draw per unit area, all while maintaining a similar power per unit area as present pixel detectors. Here we present the first measurements of the ITkPixV2 chip operated at the limits of the full chip design requirements, including in particular a measurement of the activity-induced current of the chip as a function of increasing hit rate.

Le Pottier, L↗
DOE OSTI2020

Ion beam irradiation of ABO4 compounds with the fergusonite, monazite, scheelite, and zircon structures

The effects of irradiation on CaWO4, SrWO4, BaWO4, YVO4, LaVO4, YNbO4, and LaNbO4 were investigated on thin crystals using 1.0 MeV Kr ions at 50-1000 K. All of the ABO4 compounds can be amorphized with calculated damage cross sections (σa = 1/Fc0) in the range of ~0.30-1.09 × 10-14 cm2 ion-1 at zero Kelvin. The analysis of fluence-temperature data returned critical temperatures for amorphization (Tc) of 311 ± 1, 358 ± 90, 325 ± 19, 415 ± 17, 541 ± 6, 636 ± 26, and 1012 ± 1 K, respectively, for the compounds listed above. Compared with previous in situ irradiation of ABO4 orthophosphate samples using 0.8 MeV Kr ions, the Tc values of LaVO4 and YVO4 are higher than those of LaPO4 and YPO4 by 82 K and 124 K, respectively. The Tc values of the three scheelite structures, CaWO4, SrWO4, and BaWO4, indicate that they are the most radiation tolerant compounds under these conditions. The A-B cation anti-site energies, EfAB, determined by DFT range from 2.48 to 10.58 eV and are highly correlated with the A-B cation ionic radius ratio, rA/rB, but are not correlated with Tc across the different structure types, suggesting that the formation and migration energies of Frenkel defects play a more important role in damage recovery in these compounds. We also discuss the role of cation and anion charge/iconicity as determined by DFT. ABO4 compounds with the zircon structure and B = P or V have a distinct advantage over those with B = Si as the damaged regions do not appear to be significantly affected by polymerization of (PO4) 3- or (VO4) 3- groups which might stabilize the amorphous fraction and ultimately lead to phase separation as observed in zircon (ZrSiO4).

amorphization↗
DOE OSTI2023

Signal Processing of Multiplexed Optical PWM Signals for Sensor Arrays in Nuclear Environments

Safe and effective generation of terrestrial nuclear power greatly benefits from the actionable data provided by the array of sensors located throughout a plant to provide a holistic online indication of reactor operation. This array includes sensors for monitoring coolant flow and pressure, temperature and heat transfer, radiation levels, structure health monitoring, and other critical parameters for reactor operation. While sensors capable of measuring these parameters have been developed, the electronics used in the pre-amplification and analog-to-digital conversion of the small signals they produce are extremely sensitive and susceptible to damage by high-temperatures and radiation environments nuclear reactors encounter while generating power. The small signals from sensors in nuclear power plants (NPPs) are transmitted over long cable runs which introduce dispersion artifacts into the signals of interest as well as electromagnetic interference (EMI) from lighting fixtures, pumps, mains electricity, and other equipment. To overcome these challenges, a front-end digitization (FREND) platform has been developed to use radiation-tolerant electronics to multiplex, amplify, and optically encode signals from an array of sensors for transmission over an optical fiber to mitigate dispersion and EMI artifacts from long runs of electrical cabling. To recover the optically transmitted data, a signal processing scheme based on 1-dimensional template matching to an indexing channel is described herein and demonstrated to have an effective bit-depth of 9.2 bits (1%). This scheme has been validated in proof-of-concept, non-nuclear testing and preliminary experimental results show good agreement between the measured optical output and and sensor input signals. The FREND system represents a low-loss data link between sensors in nuclear environments and data acquisition hardware which is aimed at improving the signal-to-noise ratio of the data acquired from these sensors to provide better information to operators and researchers.

Sweeney, Dan↗
DOE OSTI2020

Advanced Moderation Module for High-Temperature Micro-Reactor Applications

Development and deployment of micro-reactors that provide competitive efficiency, prevailing compactness, and inherent safety is an immediate focus of the U.S. nuclear industry. For thermal neutron micro-reactors operating at elevated temperature for optimized efficiency, such as micro molten-salt reactors (MSRs), heat-pipe reactors, and very-high temperature reactors (VHTRs), high-performance moderator based on metal hydride can enhance the neutron economy and therefore achieve reduced weight and enhanced portability. As unclad metal hydride inevitably decomposes at elevated temperature, an enclosure is required for hydride moderator to deliver desired performance at elevated temperatures. Conventional moderator enclosure solutions based on high-temperature alloys introduce extraneous neutron penalty into the micro-reactor, affecting the neutronic benefits provided by the hydride moderator. Additionally, the compatibility between the high-temperature alloy enclosure and the micro-reactor matrix is also a potential issue. In this report, we disclose an Advanced Moderator Module (AMM) concept enabled by an innovative enclosure solution combining the advantages of refractory metals, ceramic matrix composites (CMCs), and advanced coating technology to serve as hydrogen permeation barrier up to very-high temperatures. A schematic description of the AMM structure is illustrated in Figure A1. The AMM contains a moderating material core made of metal hydride with high thermal stability, such as YH 2-x . The hydride core is enclosed by an H 2 barrier layer coated on a ductile refractory metal liner to minimize hydrogen loss during high-temperature operation. A ceramic matrix composite (CMC) cladding is adopted to provide further structural strength, especially during power transients. Between the CMC cladding and metal liner, an extra diffusion barrier coating is inserted to suppress the chemical interaction at elevated temperatures. Hence, based on a series of innovative material solutions, the AMM is capable of containing the metal hydride core at elevated temperature (>900oC) inside coated and lined CMC envelop with negligible hydrogen loss throughout the microreactor lifetime. The benefit of the AMM technology was assessed based on a comprehensive multi-stage reactor-physics analysis. Advanced moderation based on hydride metals (such as YH2-x) enables reaching optimum moderation with higher fuel content than traditional VHTR technology, which is required to design compact micro-reactor cores targeting long-life operation. The AMM enclosure technology provides lower thermal neutron absorption rates than traditional solutions (based on stainless steel or Mo-based alloy such as TZM), which reduces the fissile enrichment requirements by 6-8% on a TRISO-fueled design based on the EMPIRE core. Finally, combining the hydride moderator with neutron transparent enclosure solutions provides significant potential to boost neutronics performance of micro-reactors in terms of increased core lifetime or reduced size and weight by 30-50% on a micro-reactor based on the Holos Quad technology. The progress and plans for on-going development and demonstration efforts are also discussed in this report. The current demonstration of AMM is focused on the hydrogen diffusion barrier demonstration under thermal cycling, while radiation tolerance demonstration is planned. The next step of the demonstration plan will be the assembly of miniature AMM for high-temperature testing, which can be used as a prototype for future pilot scale demonstration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Adaptive complex nanocomposite alloys for burning plasma-material interface tunability (Final Report)

The focus of this program is discovery and development of novel self-healing and adaptive materials for the PMI (plasma-material interface) envisioned for future plasma-burning extreme environments in thermonuclear fusion reactors that can provide enhanced radiation-tolerance or resistance. This program was conceived and stems from the PI’s DOE Early Career Award work on harnessing nanotechnology and mesoscale materials design in refractory metals to address gaps in PMI research. These gaps pertain to the lack of understanding of multi-scale interactions at the plasma-material interface and the development of novel material interfaces that can be designed to adapt to extreme fusion reactor conditions. The final report is separated into two primary sections: The first section consists of the program’s first two years (FY16-FY17) performance period. The second section consists of FY18 to FY19 period, where the FY19 period was a supplemental addition to support the completion of a PhD thesis and continuation of another. This work consists of an excellent team consisting of a co-PI, one postdoctoral researcher, two graduate students, two undergraduate students and collaborators both domestic and abroad.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Summary Report on the Refined User Requirements for U.S. Fusion Prototypic Neutron Source

The Fusion Prototypic Neutron Source (FPNS) is proposed as an urgent near-term facility necessary for understanding D-T fusion neutron degradation processes in materials and for the development of high-performance radiation tolerant materials for fusion power reactors. This facility is central to our goal of predicting the behavior of materials in the harsh D-T neutron environment and calibrating and verifying the materials performance models necessary to support design of next-generation fusion reactors. Key FPNS performance metrics were previously defined for the cost-efficient facility necessary to adequately and expediently resolve key materials science knowledge gaps required to move to next-step fusion devices. This brief report is intended to expand upon this previous metrics discussion as an aid to selecting the most desirable FPNS concept and enabling a facility that meets the program needs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Versatile Heat Transfer Module

Development of micro-reactors that can be easily transported to remote areas for civilian or military applications is an immediate focus of U.S. nuclear industry and therefore is being pursued by multiple reactor vendors (e.g. HolosGen LLC, Westinghouse, Ultra Safe Nuclear Corporation, URENCO, etc.) and actively supported by DOE-NE, DOE-ARPA-E and by the Department Of Defense (DoD). Very-high-temperature gas-cooled (VHTR) and heat pipe cooled reactor technologies are regarded as promising concepts for microreactors applications since they enable compact power conversion systems and display beneficial inherent safety features. Development of microreactors with enhanced performance must be supported by innovations in nuclear technologies. Several technologies being developed by the microreactor industry are being revisited with material technologies developed at the Argonne National Laboratory (ANL). Innovative high-temperature compatible solutions are developed to enable advanced thermal-spectrum microreactor designs featuring improved neutronics performance at high-temperature operating conditions to enable high thermal efficiency and economic competitiveness. This report focuses on the Versatile Heat Transfer Module (VHTM) technology that is developed at ANL for heat-transfer in thermal-spectrum microreactors. This technology is built on a liquid alkali metal heat pipe concept while leveraging advanced nuclear material technologies for high-temperature applications at reduced neutronic penalty with potential improvement in inherent safety behavior. The VHTM can be applied to design high-temperature thermal-spectrum microreactor systems with reduced size, weight, uranium enrichment, or increased lifetime. The INFINIT technology deriving from the VHTM enables design of a highly compact long-life microreactor concept for fission surface power (FSP). Demonstration of multiphysics modeling capability was also completed to showcase the added inherent safety benefits obtained with this concept. The VHTM technology remain at an early development stage with low Technology Readiness Level and Manufacturing Readiness Level. The current demonstration of these technologies has focused on manufacturing capability, on performance of the hydrogen diffusion barrier under thermal cycling, while radiation tolerance demonstration is planned. The next step of the demonstration plan will be the assembly of miniature VHTM for high-temperature testing, which can be used as a prototype for future pilot scale demonstration. Such research is critical to pursue development of these technologies that have the clear potential to significantly improve the performance of currently investigated micro-reactors systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Large-Scale Additive Manufacturing of Silicon Carbide with Process Monitoring

This report summarizes new capabilities in additive manufacturing (AM) of large-scale silicon carbide (SiC) components under the Advanced Materials and Manufacturing Technologies (AMMT) program. SiC is a promising material that is being considered for many advanced reactor designs due to its high temperature strength, radiation tolerance, minimal neutron absorption, and oxidation resistance [1-8]. One of the primary limitations to using SiC as an in-core structural material is the inability to fabricate large SiC components with complex geometries. While under the former Transformational Challenge Reactor (TCR) program, ceramic AM systems at the Oak Ridge National Laboratory (ORNL) were limited in the quantity and overall size achievable for producing ceramic AM components. To address this, ORNL has improved its infrastructure for binder jet and chemical vapor infiltration (CVI) processes with cutting-edge equipment coming online in a newly renovated laboratory space capable of producing larger components and higher throughput without compromise on part size. This report details the laboratory renovation progress, focusing specifically on the DesktopMetal (formerly ExOne) X25Pro binder jet, a large scale CVI furnace, and the added functionality of a software tool called Peregrine for quality assurance and control purposes.

36 MATERIALS SCIENCE↗
DOE OSTI2020

SAMPA Based Streaming Readout Data Acquisition Prototype

We have assembled a small-scale streaming data acquisition system based on the SAMPA front-end ASIC. The 32-channel SAMPA chip was designed for the high-luminosity upgrade of the ALICE Time Projection Chamber (TPC) detector at the CERN Large Hadron Collider. The goals of the prototype system are to determine if the SAMPA chip is appropriate for use in detector systems at Jefferson Lab, and to gain experience with the hardware and software required to deploy streaming data acquisition systems in nuclear physics experiments. The 800 channel system is composed of components used in the ALICE TPC data acquisition upgrade. Five front-end cards (FEC) support five SAMPA chips each. SAMPA data streams on an FEC are concentrated into two high-speed (4.48 Gb/s) serial data streams by a pair Gigabit Transceiver ASICs (GBTx). These ten streams (44.8 Gb/s) are transmitted from the FECs over fibers to a PCIe based readout unit. The FPGA engine on the readout unit compresses data for transmission to a server via 100 Gb ethernet. Components on the FECs are radiation tolerant. High data rates can be handled. The system is by design scalable and thus provides a functional prototype for high-rate streaming readout at Jefferson Lab and the future Electron Ion Collider. We have made fundamental measurements (noise, linearity, time resolution) on the SAMPA ASIC. We have also coupled the readout system to a small Gas Electron Multiplier (GEM) detector and have studied its response to cosmic rays. A beam test of the system is planned.

ABBOTT, David↗
DOE OSTI2022

Ion Exchange Processing of AP-105 Hanford Tank Waste through Crystalline Silicotitanate in a Staged 2- then 3-Column System (Rev.1)

The Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions LLC (WRPS), will send initial low-activity Hanford waste tank supernate feeds to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at the WTP. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B), was selected as the ion exchange media at TSCR. CST is a non-elutable inorganic material that has demonstrated robust chemical, physical, and radiation tolerance while maintaining functionality. However, exchange kinetics of Cs onto CST is slow, resulting in low utilization of the CST Cs load capacity before unacceptable Cs breakthrough. Two process flow designs have been tested, as follows. 1. Lead-lag column processing: The lead column was removed after the lag column effluent reached the waste acceptance criteria (WAC) limit, the lag column was moved into the lead position, and a new lag column was installed. This format used ~52% Cs load capacity on the lead column. 2. Lead-lag-polish column processing: The processing was stopped when the polish column effluent reached the WAC limit. This format resulted in 81% Cs load capacity on the lead column. Testing with diluted feed from Hanford tank AP-105 (AP-105DF) incorporated a nuanced change to the lead-lag-polish column system where the polish column was inserted when the lag column effluent reached WAC limit. A 10.9-L volume of AP-105DF (diluted to 5.6 M Na) was processed through the Direct Feed Test Platform system, established at Pacific Northwest National Laboratory to support small-scale waste qualification efforts. The columns consisted of 10-mL CST beds (CST Lot 2002009604, sieved to screen out >30-mesh particles) placed in 1.5-cm-inner-diameter columns. Feed was processed at 1.83 bed volumes (BV) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. The <30-mesh CST sieve cut was expected to provide appropriate performance scaling to a full-height column. The installation of the polish column later in processing (after processing 523 BVs) did not appear to fundamentally change the utilization of the lead column for Cs exchange nor did it extend the total feed processing volume when compared to the previous test with AP-107 feed. Table ES.1 and Figure ES.1 summarize the measured AP-105DF Cs load performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Cesium Removal from 5.5 and 7.0 M Na AP-105 Using Crystalline Silicotitanate

The Tank Side Cesium Removal (TSCR) system, currently operational by Washington River Protection Solutions LLC (WRPS), prepares initial low-activity Hanford waste tank supernate feeds for the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at the WTP as governed by a waste acceptance criterion (WAC). Specific to 137 Cs, this requirement is < 3.18E-5 Ci 137 Cs/mole of Na. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B), has been selected as the ion exchange media for TSCR. CST is a non-elutable inorganic material that has demonstrated robust chemical, physical, and radiation tolerance while maintaining functionality. However, testing to date on actual tank waste samples has been limited to Na concentrations between 5 and 6 M Na (Fiskum et al. 2019a, 2021b and Westesen et al. 2021a, 2021b, 2022) while actual tank conditions can reach upwards of 9 M Na. Testing with feed from Hanford tank AP-105 incorporated testing at both 5.5 and 7 M Na in order to evaluate the impact of Na concentration on volume of waste processed before reaching the WAC. A 6-L volume of 5.5 M Na AP-105 and an 8-volume of 7 M Na AP-105 was processed through the Radioactive Waste Test Platform system, established at Pacific Northwest National Laboratory to support small-scale waste qualification efforts. The columns consisted of 6- to 9-mL CST beds (CST Lot 2002009604, sieved to screen out >30-mesh particles) placed in 1.5-cm-inner-diameter columns. Feed was processed at 1.9 bed volumes (BV) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. Table S.1 and Figure S.1 summarize the measured AP-105 Cs load performance for each feed condition.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Assessment of Acoustic Sensor Application to Structural Health Monitoring of Reactor Components

The objectives of SHM of advanced fission reactors include the following: (1) Maintain safe, reliable, and efficient operation of structures, systems, and components in accordance with design intent; (3) Reduce cost; (4) Improve the comprehensive life management of structures, systems, and components; and (5) Extend the operational life of a power system through retirement for cause. Advanced reactor designs currently being considered are expected to operate at higher temperatures than light water-cooled reactors and to support missions beyond baseload electricity generation. Methods to monitor and detect for these mechanisms will require structural health monitoring (SHM) techniques, with ultrasonic methods an ideal candidate given their widespread use for NDE. This document summarized the state of technology for ultrasonic technologies, as part of an assessment of technology gaps and needed research. The great majority of sensor development to date addresses elevated temperatures and radiation tolerance, and more work is needed in both areas. Little data exists regarding corrosion effects of advanced coolants on sensor materials and couplants. Of high importance is the need for ARDs and regulatory bodies to engage and define what level of SHM is needed for autonomous or semi-autonomous control of reactors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Milestone report for MRT 8479: High Yield Xray Imager Preliminary Design Review

The High Yield Xray Imager (HYXI) is a new NIF target diagnostic system currently under development. The goal of HYXI is to provide high-fidelity, high temporal resolution x-ray imaging capability on high yield NIF implosions at 10MJ and above. The HYXI instrument design concept is based on the combination of two technologies that have been successfully utilized at the NIF on previous instruments, electron pulse-dilation and hybrid-CMOS sensor imaging. The combination of these two techniques will give HYXI sufficient data quality to ascertain differences in hot spot formation dynamics between high and low yield implosions. This information will highlight the critical hot spot conditions needed for ignition and burn. The HYXI design leverages the successful operation of the PDIXI x-ray imager at the NIF on multi-MJ yield shots. Also, a new radiation tolerance CMOS imaging array is being developed to eliminate the significant background noise which limits the data quality of PDIXI. The HYXI Preliminary Design Review was completed at the end of Q4 FY23. The HYXI project is a multi-year effort with a phased approach to be bring up system functionality over time in parallel with the development and fabrication effort of the new CMOS imaging array. Initial time integrated NIF data will be collected in Q2 FY25, first time resolved imaging with Daedalus starting in Q2 FY26 and the final performance qualification of the complete HYXI system is scheduled for Q2 FY27.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Development Plan for Acoustic Emission and Vibration Sensors

This document discusses the need for high temperature, radiation tolerant sensors for monitoring vibration and acoustic emissions in advanced reactors and irradiation experiments. Testing of several commercial sensors using infrastructure available at Idaho National Laboratory was performed to benchmark currently available sensors as well as the test facilities. Finally, a strategy for developing sensors for more extreme environments is presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Oxide Dispersion Strengthened Ferritic Steel Wire Feedstock Development for Large Format Additive Manufacturing - CRADA 620 (Abstract)

This CRADA project funded through DOE’s INFUSE program seeks to demonstrate the viability of fabricating large, complex parts from oxide dispersion strengthened (ODS) steel with advanced manufacturing. Exhibiting excellent radiation tolerance and high mechanical performance at elevated temperatures, ODS steel is a promising structural material candidate for near-plasma components in fusion energy systems. Its use, however, has been limited by a lack of manufacturability. This project will seek to produce ODS steel wire through a solid-state shear assisted extrusion process and then demonstrate that the wire can undergo controlled local melting while being welded with the final part sufficiently retaining the beneficial properties of ODS steel. This would allow the use of wire-arc additive manufacturing (WAAM) to manufacture large-scale ODS parts, even though ODS is currently only available as a powder. WAAM is a promising technique for producing components like the replaceable ARC vacuum vessel in CFS’ fusion reactor design. Meanwhile, this project will also expand PNNL’s capability in producing custom wire feedstock with friction extrusion, enabling downstream large-scale manufacturing with WAAM and solid-state based additive manufacturing.

36 MATERIALS SCIENCE↗