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DOE OSTI2020

Demonstrate embedding of sensors in a relevant microreactor component

This report describes the first successful embedding of sensors in stainless steel pipe specimens and a miniature hex block to support structural health monitoring of microreactor components during electrically heated testing in non-nuclear testbed facilities. Sensors were embedded with an ultrasonic additive manufacturing process that uses a combination of downward pressure and an ultrasonic scrubbing motion to bond stainless steel foils to underlying materials. After placing sensors within cavities that have been machined to the size of the sensors, the ultrasonic bonding process fills the gaps around the sensors and embeds them within the part. Both Type K thermocouples and spatially distributed fiber-optic strain sensors were successfully embedded within two pipe specimens and a miniature hex block. After the fiber-optic sensors were embedded, they were interrogated and found to have minimal signal attenuation within the embedded region—a significant improvement compared to previous trials. This is likely because of the low-bend-loss fiber that was used, which greatly reduces microbending losses that occur after the embedding process. In addition to showing minimal signal attention, the fibers also showed adequate compressive strain, which is a strong indication that the sensors are physically embedded in the parts as opposed to simply being loosely contained inside the machined cavities. The embedded fibers were able to capture the strain induced in one of the pipe specimens while the pipe was actively heated with a moving heat source.

42 ENGINEERING↗

DOE OSTI2020

Transportation analysis and related design optimization of the Fermilab high-beta 650 MHz cryomodule

The Proton Improvement Plan-II (PIP-II) at Fermi National Accelerator Laboratory (FNAL) will create a new and vastly improved accelerator, which will be the source of high-energy particles for the experiments taking place at FNAL. The new linear accelerator (LINAC) contains several types of cryomodules, which are individual particle accelerators. The last cryomodule in the LINAC will be the High-Beta 650 (HB650), which will operate at 650 MHz. Each module is approximately 15 meters in length and 1.5 meters wide, weighs 13 tonnes, and shares many design features with three of the other cryomodules. The HB650 consists of two primary sections, the outer vacuum vessel and the components that reside within it, which will be cooled to cryogenic temperatures – the cold mass. Once assembly is complete at FNAL, it will be transported to one of three places: another location at FNAL, another national laboratory within the continental United States, or to a scientific partner in Europe. Any excitation the module experiences, such as going over rough roads when on a semitrailer, can create high stresses in components and cause failure if severe enough. Additionally, any delicate components with low resonant frequencies that match the excitation spectrum could achieve resonance, potentially causing a fatigue failure by repeated flexing. To ensure the successful transportation of the HB650, a transport analysis utilizing ANSYS simulation software has been performed on major subassemblies, as well as analysis of the subassemblies as a combined system. The optimization of components, design of support structures, and overall increasing of the lowest resonant frequencies have resulted in a cryomodule that will be better equipped to handle transportation and any large impact loads that come with it.

43 PARTICLE ACCELERATORS↗

DOE OSTI2021

Summary of US DOE R&D Activities on Graphite Oxidation (2006–2021)

The objective of the international collaboration between United States Department of Energy (U.S.-DOE) and Generation IV International Forum (GIF) is the development of the next generation of nuclear energy systems. The current GIF Project Arrangement (PA) on Materials (2018-2022) was revised in 2019 and extended for another 10 years (2020-2030). The Work Package 1 (“Graphite”) of the extended Project Plan (PP) on Materials specifies technical tasks and High Level Deliverables for research and development (R&D) activities related to using graphite in fuel elements, reflectors, and support structures of Very High Temperature Reactors (VHTR). The graphite tasks include specification and acquisition, qualification and development of new grades, characterization of properties, and development of behavior models. Specifically, Task 1.4 (“Graphite Oxidation Behavior”) outlines planned activities related to acute oxidation by air and chronic oxidation by impurities in the helium coolant. A final report on experimental data regarding graphite oxidation behavior is scheduled for 2022 (deliverable 3.1.1.4.a). In preparation of this deliverable, this document summarizes the R&D activities funded by U.S.-DOE from 2006 (the inception of the VHTR system arrangement) through present (2021). This report is being submitted to the GIF Graphite Working Group (GWG) to serve as input for the GWG high-level deliverable to the Project Management Board (PMB) of PA on Materials. Besides U.S.-DOE, other organizations participating to Task 1.4 of the current PA on Materials are: European Commission’s Joint Research Center (JRC), Korea Atomic Energy Research Institute (KAERI), and Japan Atomic Energy Agency (JAEA). U.S.-DOE is the main contributor on graphite oxidation R&D, with 85 % commitment of total funding during 2018-2022.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Flow Sensor Test Article (F-STAr) - Status Report for FY2021

This report will review updates on the status of F-STAr. First, the report will describe a general overview of the test article and how each component interacts. Then, each main component will be discussed in a deep dive section which will review the design and other relevant details. F-STAr components which are discussed in more detail include the 120 GPM-pump, the support structure and test sections, 5 kW-heater, the 2 kW-cooler and cooling system, and the submersible flowmeter. Finally, this report will conclude with a brief recap and a bulleted outline of the path moving forward.

42 ENGINEERING↗

DOE OSTI2021

MRSt S Specifications Report

The MRSt spectrometer is a magnetic spectrometer designed for momentum analyzing deuterons (10.7 – 14.2 MeV). The MRSt spectrometer will be attached to the target chamber of the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNS), Livermore, California. This document defines the technical requirements for the design, analysis, fabrication, inspection, testing and delivery of the complete set of magnets, vacuum chambers, and accessories to become part of the spectrometer. The interface issues, including the support structures, need to be defined later by LLNS and the vendor, which will require time and dialog before a solution can be found. This “Specification Report for the MRSt-S System”, January 19, 2020 is based on and supersedes an earlier version “Specification Report for the MRSt-S System”, May 1, 2019 The vendor shall check these specifications for possible conflicts between statements in this document. Possible solutions to any such conflicts should be included in the proposal submitted by the vendor. The proposal shall also include the delivery FOB destination LLNS.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2021

The Role of Nuclear-Conventional Intermingling on State Decision-making and the Risk of Inadvertent Escalation

The questions answered by this report are: What are the implications of nuclear and conventional intermingling on crisis stability and the potential risk of miscalculation?, and Specifically, how might entanglement influence US and competitor decision making during crisis and conflict? In practice, there are three main forms of nuclear-conventional intermingling. First, intermingling between nuclear and non-nuclear weapon systems can occur via the fielding of dual capable delivery systems like missiles or aircraft. Second, intermingling can happen due to the co-location of nuclear and non-nuclear forces and their support structures—for example, the co-location of strategic bombers and general-purpose aircraft, or the co-location of strategic submarines and general-purpose vessels. Third, intermingling can occur via convolving nuclear and conventional military command and control systems, to include ballistic missile early warning and potentially space surveillance systems as well. All three forms of nuclear-conventional intermingling have significantly increased since the end of the Cold War, driven by both technological and doctrinal changes. However, there are important differences in the rationale behind, and also the risks associated with these three different forms of intermingling. The mere existence of dual capable systems is not new— deploying such systems can increase the effectiveness of forces, and it can also provide more flexibility. The major powers have both employed and threatened with dual capable systems for decades, and they have done so without nuclear escalation. Similarly, the major powers co-located nuclear and conventional systems in the Cold War, and they did so for variety of reasons that had nothing to do with complicating the adversary’s risk calculus. The Soviet Union, for example, decided to co-locate its nuclear and conventional forces for economic and administrative reasons. Although today it might be recognized as a useful deterrent tool, it was not their primary intention. Major powers want to convince rivals that the co-location of forces creates a high bar for targeting and raises the risk of nuclear escalation, but they also want to have the flexibility of this not being true in an actual crisis or conflict.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2021

Report on High Energy Arcing Fault Experiments: Experimental Results from Open Box Enclosures

This report documents an experimental program designed to investigate High Energy Arcing Fault (HEAF) phenomena. The experiments focus on providing data to better characterize the arc to improve the prediction of arc energy emitted during a HEAF event. An open box experiment allow for direct observation of the arc, which allows diagnostic instrumentation to record the phenomenological data needed for better characterization of the arc energy source term. The data collected supports characterization of the arc and arc jet, enclosure breach, material loss, and electrical properties. These results will be used to better characterizing the hazard for improvements in fire probabilistic risk assessment (PRA) realism. The experiments were performed at KEMA Labs located in Chalfont, Pennsylvania. The experimental design, setup, and execution were completed by staff from the NRC, the National Institute of Standards and Technology (NIST), Sandia National Laboratories (SNL) and KEMA Labs. In addition, representatives from the Electric Power Research Institute (EPRI) observed some of the experimental setup and execution. The HEAF experiments were performed between August 22, 2020 and September 18, 2020 on near-identical 51 cm (20 in) cube metal boxes suspended from a Unistrut support structure. The three-phase arcing fault was initiated at the ends of the conductors oriented vertically and located at the center of the box. Either aluminum or copper conductors were used for the conductors. The low-voltage experiments used 1 000 volts AC, while the medium-voltage experiments used 6 900 volts AC consistent with other recently completed experiments. Durations of the experiment ranged from 1 s to 5 s with fault currents ranging from 1 kA to 30 kA. Real-time electrical operating conditions, including voltage, current and frequency, were measured during the experiments. Heat fluxes and incident energies were measured with plate thermometers, radiometers, and slug calorimeters at various locations around the electrical enclosures. The experiments were documented with normal and high-speed videography, infrared imaging and photography.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

US HL-LHC Accelerator Upgrade Project (MQXFA Final Design Report)

The MQXFA Quadrupole magnets will be installed in High Luminosity LHC to form the Q1 and Q3 inner triplet optical elements in front of the interaction points 1 (ATLAS) and 5 (CMS). A pair of MQXFA units is assembled in a stainless steel helium vessel, including the end domes, to make the Q1 Cold Mass or the Q3 Cold Mass. The US HL LHC Accelerator Upgrade Project* is responsible for the design, manufacturing and test of the Q1/Q3 Cold Masses and the complete MQXFA magnets. CERN provides the cryostat components and is responsible for integration and installation in HL LHC. The MQXFA quadrupoles have 150 mm aperture, 4.2 m magnetic length, nominal gradient of 132.2 T/m, and coil peak field of 11.3 T. They use Nb_3Sn conductor and a support structure made of segmented aluminum shells pre-loaded by using bladders and keys. This report presents the final design of the MQXFA quadrupole magnets.

43 PARTICLE ACCELERATORS↗

DOE OSTI2022

FY22 Progress on Computational Modeling of the Water-Based NSTF

This report summarizes the system-level modeling effort by Argonne National Laboratory (Argonne) of the Natural convection Shutdown heat removal Test Facility (NSTF) in FY22. As an extension of the effort from FY21, this year’s work focuses primarily on the two-phase modeling of the NSTF using RELAP5-3D, particularly with the inclusion of the cavity model. The results from simulations were used to compare against experimental data for benchmarking purposes of the RELAP5 deck. Additionally, RELAP5 was used as a predictive tool to guide planned test operations and identify expected system behaviors. In the first part of this report, details are provided of the cavity omitted model where heat flux is applied directly as a boundary condition to the risers. The general trend predicted by the RELAP5 model matches that from the experimental data when a single-phase natural circulation flow is first established, followed by an oscillatory two-phase period and finally a stable two-phase flow. However, the onset of oscillations is predicted early by the model due to the smaller thermal mixing region in the tank. However, by expanding the simulated thermal mixing region in the tank, the onset of oscillations predicted by the model is able to match that from the experiment. These oscillations where studied in depth and are deduced to be flashing-induced instability. The model was then modified to simulate an accident scenario case where a representative heat load based on the full-scale Framatome’s 625 MW t SC-HTGR was applied directly to the riser channels. The simulated initial and boundary conditions were identical to those performed experimentally, facilitating direct comparisons between the predicted and experimental results. It was determined that the results showed some discrepancies remain, likely due to the overprediction of vapor generation rate by the computer model. In the second part of this report, the cavity model is re-introduced where it is observed that the RELAP5 prediction is now able to capture the major trends of the observed flow commonly observed during two-phase conditions. However, the onset of oscillations is once again predicted early by the model, possibly caused by the underprediction of heat loss from the heater and cavity. This is likely due to the omission of support structures in the cavity that can act as additional pathways for heat to escape to the environment. To overcome the underprediction of heat loss, part of the insulation surrounding the cavity side panels and the back of the heaters are removed to allow heat to escape directly to the environment, which then improves the RELAP5 prediction. Parametric studies are also performed to investigate the effects of heater power, tank inventory level, and tank gas space pressure on flow behaviors, also in direct comparison to conditions tested experimentally. User option-61 in the RELAP5-3D input deck, which changes the heat transfer coefficient correlations used for calculating the vapor generation, is also investigated where it is found that by enabling the option, the overall duration of oscillations is increased and matches that from the experiment better. The RELAP5 model is further benchmarked with a header inlet- throttling case where it is observed that the prediction from the model fails to capture some major features observed in the experiment. By using a modified loss coefficient curve for the valve, the accuracy of the prediction is improved where most of the major features observed in the experiment are predicted by the model. Lastly, the model is benchmarked with an inventory depletion scenario where it is observed that despite the modeling limitation of RELAP5, the prediction shows good agreement with the experimental data where major trends and features are captured by the model. Future work will see continued development of the current RELAP5-3D input deck of the NSTF to both improve the accuracy of the model’s predictive capability and continuing serving the experimental program. The mutually beneficial relationship between analysis and experimental efforts has become integral to the parent NSTF program, and the greater objective to fully understand and accurately predict the heat removal performance of a full scale RCCS concept.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Flow Sensor Test Article (F-STAr) (Design and Fabrication Status Report)

The Flow Sensor Test Article (F-STAr) is a test article currently under construction for the Mechanisms Engineering Test Loop (METL). F-STAr was designed to provide high sodium flowrate capabilities for sensor calibration, component testing, and fluid studies. Figure 1 shows two solid model views of the new test article. F-STAr includes a high-capacity pump that can provide a nominal flowrate of 120 GPM; a test section support structure that can accommodate a wide array of sub-test articles and their instrumentation; and finally, a heating and cooling system to aid in controlling the testing environment. Initially, F-STAr will be configured to test liquid metal flow sensors, specifically field shift sensors like the Eddy Current Flow Sensors (ECFS) based on the RDT C4-7T standard. To test these sensors, two test sections were designed that attempt to model Sodium Fast Reactor (SFR) outlet conditions. Figure 2 shows models of the “Full-Scaled Test Section” (FSTS) that represents a generic SFR fuel handling socket and a “Pseudo-Scaled Test Section” (PSTS) that represents a generic array of scaled fuel handling sockets. While F-STAr will be configured to test ECFS’s, it can also be outfitted to meet other experimental needs. For example, F-STAr could be setup with a test section that includes a fluidic diode or a component test investigating the performance of hydrodynamic bearings. Other test sections include studies of sodium thermal hydraulics like thermal striping. Overall, F-STAr is a flexible test article designed to accommodate many needs. This report will provide a status update on the design and construction of F-STAr. Manufacturing of all components has commenced, and several components have already been completed. These components include the pump and test section assembly stand. Other components, such as the heater, have been completed but were rejected due to the vendor not meeting the requirements of the purchase order. Lastly, the submersible flowmeter and main flange components are under construction, and their status will be reviewed in this report. Overall, most of the F-STAr components and parts will be completed by the end of September 2022.

42 ENGINEERING↗

DOE OSTI2022

RRFMC Drop Tester: Conduct of R&D Engineering System Risk Assessment

This report provides a documented risk assessment and subsequent determination of engineering rigor for the modification of the drop tester system inside of the Respirable Release Fraction Measurement Chamber (RRFMC) located in TA03-130. This risk assessment does not apply to the modification of the structural support system, containment barrier, pressure system, nor fluid system of the RRFMC, but only seeks to address the risk associated with the specific system used to raise and drop test articles.

42 ENGINEERING↗

DOE OSTI2022

Fabrication and analysis of hydroxide/ceramic dual-phase (HCDP) membranes

Ceramic materials show potential to serve as support structures for membranes that can be used to uptake carbon dioxide from high temperature environments, like those found in nuclear flue plants. The porosity of ceramic materials allows for the infilling of those pores with molten hydroxides, that show a natural affinity for absorbing CO 2 from its surroundings. The membranes are easy to produce and were made with thickness of 0.5mm, 1.0mm, and 2.0mm. The membranes were exposed to various conditions after the infilling process and their responses were measured. The percent of hydroxides were measured after the infilled membranes had been in a tube furnace at 400°C, followed by exposure to 30 minutes of CO 2 , and lastly 4 hours of steam. Those tests yielded mole percent hydroxide results of 88.4% ± 4.6%, 68.6% ± 5.1%, and 80.7% ± 4.6%, respectively.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Investigation of Coincidence Counting for Improving Minimal Detectable Activity of 110m Ag in Single Particle Gamma Analysis

Post-irradiation examination (PIE) of fuel particles from the fourth Advanced Gas Reactor Fuel Development and Qualification (AGR) Program irradiation (AGR-5/6/7) is being performed at Oak Ridge National Laboratory (ORNL). Tristructural isotropic (TRISO)-coated particles and associated compacts for the AGR-5/6/7 experiment fabricated by BWX Technologies Nuclear Operations Group were formed into a graphite matrix compact and irradiated at the Advanced Test Reactor at Idaho National Laboratory. At ORNL, particles are deconsolidated from the graphite matrix compact and individually scanned for emitted gamma rays with the Irradiated Microsphere Gamma Analyzer (IMGA). The IMGA system comprises a single high purity germanium (HPGe) detector, an automated particle handling vacuum system, and an ORTEC DSPEC-50 digital spectrometer for gamma ray analysis. IMGA quantifies gamma ray-emitting fission product inventories of individual TRISO particles, and these inventories can be compared with the measured average inventories per particle and radionuclide inventories predicted by AGR-5/6/7 physics calculations to determine if a particle experienced radionuclide release. Details on IMGA data collection methods can be found in the literature. The TRISO particle’s SiC layer provides structural support, as well as a barrier for fission product release during irradiation or subsequent safety testing. A weakened or compromised SiC layer can be identified by the release of radionuclides, such as 137 Cs, which is detected by IMGA. However, select radionuclides, such as 90 Sr, 110 mAg, and 154 Eu have been shown to migrate through an intact SiC layer. Measurement of the radionuclide 110 mAg is significant as its release has been shown to be particularly sensitive to in-reactor conditions (e.g., temperature) with broad variable particle to-particle release behaviors observed within a single compact. As such, 110 mAg activity is often used for particle selection for comprehensive PIE as bounding 110 mAg retention particles are hypothesized to represent limits in particle behaviors within a compact. As TRISO particle fuel PIE activities continue over time, IMGA measurements of the 110 mAg inventory are eventually hindered because of its relatively short half-life (~250 days). As the fuel ages from its end of irradiation (EOI) date, the measurement uncertainty and minimum detectable activity (MDA) of 110m Ag increase because the detector background continuum begins to dominate. For particles from the second AGR irradiation experiment (AGR-2), the 110m Ag MDA was above 20% of the calculated average particle inventory after approximately five half-lives, and 110m Ag activity was no longer measurable with IMGA after approximately seven half-lives. Therefore, coincidence counting approaches have been explored to determine feasibility of leveraging new approaches to overcome limitations associated with increasing MDA over time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Niowave Neutron Source Converter: Lead-Bismuth-Eutectic (LBE) Windowless Target Design and Evaluation

Los Alamos National Laboratory (LANL) is working with Niowave on the design and evaluation of their lead-bismuth-eutectic (LBE) windowless target (i.e., neutron source converter). Niowave plans to use 200 kW electron beam at 40 MeV beam energy to produce neutrons by photonuclear reaction with LBE. Then, the neutrons undergo fission at the surrounding uranium target assembly (UTA) to produce Molybdenum 99 (Mo-99) as a fission product, which eventually decays to Technetium-99 (Tc-99m). Tc-99m is one of important radioisotopes that is used for medical diagnostics. LANL conducted 3D multiphysics analysis for the Niowave neutron converter design and provided design assessment in thermohydraulic aspects. LANL conducted radiation transport calculations using Monte-Carlo N-Particle (MCNP) code with unstructured meshing scheme. The 3D volumetric heating profiles in the LBE and Stainless-Steel (SS) housing were imported into multiphase computational fluid dynamics (CFD) to obtain 3D temperature profiles of LBE and SS through conjugate heat transfer (CHT) analysis. The key findings are: LBE film thickness at the center of the beam is approximately 1.6 cm with a maximum LBE velocity of approximately 1.8 m/s, which is below a 2 m/s limit to avoid erosion issues on supporting structures; Heat deposition in the LBE peaks at ~1 cm depth from the LBE free-surface because of the forward interactions of electron, photon, and neutron with LBE; LBE maximum temperature is ~360 °C, which is below LBE evaporation initiative temperature, ~450 °C; LBE-SS interface temperature is ~350 °C, which is below the safety thermal limit to prevent severe corrosion on SS. The results indicate that Niowave’s neutron converter design satisfies both hydraulic and thermal criteria for safe operation. By virtue of such computational analysis, Niowave can move toward establishing an experimental setup to experimentally test their LBE neutron converter. The following sections describe the detailed work done by LANL.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 (Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. To date a test matrix has been completed utilizing the primary system of THETA. These tests, along with computational fluid dynamics and systems code models, determined the heat transfer across the core barrel and intermediate heat exchanger outlet was too great to effectively represent scaled thermal hydraulic phenomena of a liquid metal cooled reactor. Therefore, a significant effort was made to remove the primary system from the METL 28” test vessel #4 and clean the residual sodium from the primary system to facilitate upgrades. Thermal insulation was then incorporated in the core barrel and intermediate heat exchanger outlets. The primary system was then replaced, and a series of tests were performed to assess the performance of the thermal insulation. With primary system testing and upgrades complete, the secondary system could then be brought online. The tube side of the shell-and-tube intermediate heat exchanger was installed onto the primary system flange to begin installing the secondary system. The support structure for the secondary system was then erected on the METL mezzanine alongside the THETA primary system to facilitate installation of the secondary system components (sodium-to-air heat exchanger, flowmeter and pump). The piping and expansion tank were welded into the secondary system. Non-destructive examination of the secondary system welds was completed in order to satisfy ASME B31.3 pipe code for class M process fluids. The heating system and insulation were then added to prepare the system to be filled with sodium. The ancillary electrical equipment was installed which included the pump control box, blower VFD, pipe heater control system, etc. The secondary system will be filled, and a test matrix will be completed in early FY2024.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Building the next high voltage dc photogun polarized source for the ILC at JLab

During the ILC Global Design Effort (2007) two 200 kV high voltage photoguns based on an inverted insulator geometry were developed in 2010 and successfully tested at Jefferson Lab (JLab). One of the photoguns has been operated at 130 kV in the CEBAF accelerator for the JLab nuclear physics program. The inverted insulator replaces the electrode support structure, consequently significantly less metal prone to field emission is biased and surfaces contributing to the vacuum load are significantly reduced. Implementing a biased anode, and using a larger laser spot size has made this the state of the art polarized beam source with ~200 C charge lifetime from SSL GaAs photocathodes. However, operating at the design 200 kV voltage has been hindered by field emission at ~190 kV. To reliably operate without field emission requires lengthy conditioning applying 50-100 kV beyond the voltage needed for beam operations, without breakdown or damage to the insulator. Over the past decade, JLab has developed two photoguns with a larger version of the commercial inverted insulators used in the ILC design. One photogun operated at 300 kV for magnetized beam studies, and the concept has been implemented in the BNL polarized photogun. The second photogun has recently been installed in CEBAF to operate at 200 kV field emission free. Higher voltage photoguns help the ILC and CLIC achieve higher bunch charge (2013 TDR), and additionally provide margin for operating with shorter laser pulse length and/or smaller laser spot size (better emittance at higher peak current). Such photogun designs need to reach ~400 kV without breakdown or insulator damage to operate field emission free at > 300 kV. This also allows for direct injection into a SRF booster. JLab submitted a proposal to HEP for developing in partnership with industry an inverted geometry insulator to fit commercial 400 kV cables. This contribution describes JLab experience developing and testing high voltage photoguns, and what is needed to sufficiently condition for voltage levels higher than beam operational voltage.

Hernandez-Garcia, Carlos↗

DOE OSTI2022

Task 12 PV Sustainability - Mineral Resource Use Footprints of Residential PV Systems

Resource use intensity is often mentioned as one of the main characteristics of PV systems and PV electricity. Recently, the International Energy Agency published a report on the role of critical minerals in clean energy transitions. The Product Environmental Footprint pilot study on PV electricity quantified (among other environmental impacts) its abiotic depletion potential. So far, a comprehensive assessment of resource use impacts highlighting the different facets of its impacts is however lacking. For the first time, the resource use impacts of PV electricity are quantified simultaneously with four impact category indicators recommended or suggested by the Life Cycle Initiative hosted at UN Environment. The indicators cover distinctly different aspects of resource use, namely resource depletion with the Abiotic Depletion Potential, ultimate reserves (ADP UR ), economic resource scarcity with the Abiotic Depletion Potential, economic reserves (ADP ER ), resource quality with the Surplus Ore Potential, Ultimate Recoverable Resources (SOP URR ) and re-source criticality with the ESSENZ method. The resource use impacts caused from the generation of 1 kWh electricity with a residential scale photovoltaic (PV) system installed in Central Europe using mono- and multi-crystalline silicon panels and CdTe panels, respectively are quantified. The product system includes manufacture, use and end of life treatment (take back and recycling) of the PV panels, cabling, inverter and supporting structure, the supply chains of the raw materials and energy used in PV panel and inverter manufacture as well as transport logistics.

14 SOLAR ENERGY↗

DOE OSTI2022

Management of the Three Mile Island Unit 2 Accident Corium and Severely Damaged Fuel Debris (Rev. 2)

The Three Mile Island, Unit Two (TMI-2) pressurized water reactor core underwent a significant meltdown in 1979 due to an untimely combination of maintenance problems that led to a loss of feedwater, followed by a series of operational misunderstandings and errors. Primary coolant discharging through a malfunctioning valve represented what was analyzed as a “small-break” loss-of-coolant accident (LOCA) and ultimately became a full core meltdown. The melted core recovery process required the development of a wide array of tools. After approximately three years of water management and other cleanup actions, the first views of the core revealed a much higher degree of damage than previously expected. Approximately 62 metric tons of the core had melted, leaving only 42 of the 177 fuel assemblies standing with fuel rods intact. The core to be recovered was composed of loose, gravel-like and granular particulate material and a central solidified mass of formerly molten fuel. Molten fuel had also penetrated some of the pressure vessel internals and resolidified below the main core support structure. Robotic tools that had been designed for the task were of limited effectiveness due to the range of material types and phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗