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DOE OSTI2015

ASME Code Considerations for the Compact Heat Exchanger

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a working fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

KLYNAC: Compact linear accelerator with integrated power supply

Accelerators and accelerator-based light sources have a wide range of applications in science, engineering technology and medicine. Today the scienti c community is working towards improving the quality of the accelerated beam and its parameters while trying to develop technology for reducing accelerator size. This work describes a design of a compact linear accelerator (linac) prototype, resonant Klynac device, which is a combined linear accelerator and its power supply - klystron. The intended purpose of a Klynac device is to provide a compact and inexpensive alternative to a conventional 1 to 6 MeV accelerator, which typically requires a separate RF source, an accelerator itself and all the associated hardware. Because the Klynac is a single structure, it has the potential to be much less sensitive to temperature variations than a system with separate klystron and linac. We start by introducing a simpli ed theoretical model for a Klynac device. We then demonstrate how a prototype is designed step-by-step using particle-in-cell simulation studies for mono- resonant and bi-resonant structures. Finally, we discuss design options from a stability point of view and required input power as well as behavior of competing modes for the actual built device.

43 PARTICLE ACCELERATORS↗

DOE OSTI2018

Confirmatory LBL Analysis of AGR-5/6/7 Compacts and Over-Coated Particles BWXT Batches 11034, 14154C, 14156C, AND 14156D

Fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) was produced by BWX Technologies (BWXT) Nuclear Operations Group in Lynchburg, Virginia. Tristructural isotropic (TRISO) coatings were deposited using a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace on 425-μm-nominal-diameter spherical kernels from Lot J52R-16-69317 containing a mixture of 15.5% 235U low-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. TRISO-coated particle Lot J52R-16-98005 was over-coated with a graphite/resin blend and these over-coated particles were pressed in half-inch-diameter, one-inch-long cylindrical compacts. Two packing fractions (PF) were produced, 40%PF and 25%PF, where the TRISO particle volume made up approximately 40% and 25% of the total compact volume, respectively. The AGR-5/6/7 Fuel Specification, SPC-1352 [Marshall 2016], provides the requirements necessary for acceptance of the fuel manufactured for the AGR-5/6/7 irradiation test. Quality control (QC) acceptance testing for all AGR-5/6/7 composited lots and single batches was performed by BWXT, with the exception of pyrolytic carbon (PyC) anisotropy and defective IPyC fraction in the TRISO candidate batches and final composite, which were measured at the Oak Ridge National Laboratory (ORNL) and reported in ORNL/TM-2017/036 [Hunn et al. 2017] and ORNL/TM-2017/037 [Helmreich et al. 2017a].

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Compact Heat Exchanger Design and Testing for Advanced Reactors and Advanced Power Cycles

The goal of the proposed research is to demonstrate the thermal hydraulic performance of innovative surface geometries in compact heat exchangers used as intermediate heat exchangers (IHXs) and recuperators for the supercritical carbon dioxide (s-CO 2 ) Brayton cycle. Printed-circuit heat exchangers (PCHEs) are the primary compact heat exchangers of interest. The overall objectives are: To develop optimized PCHE designs for different working fluid combinations including helium to s-CO 2 , liquid salt to s-CO 2 , sodium to s-CO 2 , and liquid salt to helium; To experimentally and numerically investigate thermal performance, thermal stress and failure mechanism of PCHEs under various transients; and To study diffusion bonding techniques for elevated-temperature alloys and examine post-test material integrity of the PCHEs. The project objectives were accomplished by defining and executing five different tasks corresponding to these specific objectives. The first task involved a thorough literature review and a selection of IHX candidates with different surface geometries as well as a summary of prototypic operational conditions. The second task involved optimization of PCHE design with numerical analyses of thermal-hydraulic performances and mechanical integrity. The subsequent task dealt with the development of testing facilities and engineering design of PCHE to be tested in s-CO 2 fluid conditions. The next task involved experimental investigation and validation of the thermal-hydraulic performances and thermal stress distribution of prototype PCHEs manufactured with particular surface geometries. The last task involved an investigation of diffusion bonding process and posttest destructive testing to validate mechanical design methods adopted in the design process. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed s-CO 2 test loop (STL) facility and s-CO 2 test facility at University of Wisconsin – Madison (UW).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

PART 2: Initial Beam Dynamics Results of an Integrated and Compact 1.3 GHZ Superconducting Radio Frequency (SRF) Electron Source

This paper is the continuation of a technical paper which is called “Detailed Electromagnetic Design of an Integrated and Compact 1.3 GHz Superconducting Radio Frequency (SRF) Electron Source” [1]. In the previous paper, we examined if a promising RF electromagnetic design of the gun and cathode region could be achieved. Further, we examined if this gun region could be fully integrated, in terms of the electromagnetic RF design, with the rest of the accelerator, keeping in mind the overall goal of compactness and reduction of duplicative infrastructure commonly associated with separated function electron guns and structures. While we convinced ourselves thus far that the RF electromagnetic design of our concept is feasible, we also needed to perform particle tracking to assess the beam properties through the entire system, understand and remediate any beam loss. For this step, we examined the cathode region and analyzed the gating process of the thermionic cathode, as losses would most likely be attributed to releasing electrons into the accelerating structure during an unfavorable phase window.

43 PARTICLE ACCELERATORS↗

DOE OSTI2019

SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) facility. The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Originally aiming at supporting sodium fast reactor (SFR) design, it is now extended to cover other non-light water reactor (non-LWR) concepts, such as high-temperature gas-cooled reactor (HTGR) designs, molten salt-fueled, and molten salt-cooled reactor designs. To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments relevant to these reactor concepts. The Compact Integral Effects Test (CIET) facility, which was designed to reproduce the thermal-hydraulics response of fluoride salt-cooled high-temperature reactors (FHR) under both forced- and natural-circulation conditions, has been identified and selected as one of benchmark test facilities for SAM code validation. Among available CIET experiments, three sets of them with distinctive characteristics were selected for SAM code validation purpose. The three sets experiments are: 1) Power step change transient tests; 2) DHX-DRACS natural circulation tests; and 3) heater frequency response tests. For all tests, steady-state or transient, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Safety Testing and Destructive Examination of AGR-2 UO 2 Compact 3-1-1

Post-irradiation examination and elevated-temperature safety testing are being performed on compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). The compacts in the AGR-2 irradiation experiment held either tristructural isotropic (TRISO)-coated particles containing uranium oxide fuel kernels (UO 2 ) or TRISO-coated particles containing fuel kernels with both uranium carbide and uranium oxide phases (UCO). In UO 2 TRISO particles, oxygen released by uranium fission can react with the surrounding carbon in the buffer layer to form carbon monoxide (CO). Excess CO can lead to various irradiation performance issues under certain operating conditions, such as pressure-induced fracture, kernel migration, and silicon carbide (SiC) corrosion. In UCO TRISO particles, CO formation is reduced because the chemical potential for oxidation of uranium carbide is lower than for oxidation of carbon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

PIE on Safety-tested AGR-1 Compacts 5-3-3, 5-1-3, and 3-2-3

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy in support of Advanced Reactor Technologies (ART). AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE recently completed at ORNL on irradiated AGR-1 Compacts 5-3-3, 5-1-3, and 3-2-3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compact 1-4 and 10-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a “fuel body”, in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Sparse Dense and Compact Linearizations of the AC OPF (Online Supplement)

This document is an online supplement for Sparse, Dense, and Compact Linearizations of the AC OPF. Here we present complete derivations of the formulations examined, details of the lazy constraint algorithm, and full computational results supporting Sparse, Dense, and Compact Linearizations of the AC OPF.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2021

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 8-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both to be focused on obtaining fission product transport data to support improvements in modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These goals were combined in the AGR-3/4 irradiation experiment, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body, in which the outer graphite ring included a floor and cap to fully enclose the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to obtain data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Matrix Graphite Material Models In Pebbles and Compacts For Bison

The cores and reflectors in high-temperature gas-cooled reactors (HTGRs) are made of graphite materials, with the graphite acting as a moderator, a fuel host matrix, or the foundation for various structural components. This study aims to survey the models in the literature for graphite materials being used as host matrices in pebble/fuel compacts and to implement those surveyed models into Bison to conduct an early assessment of graphite's thermo-mechanical response under various reactor conditions. In this study, thermal (e.g., thermal conductivity, and specific heat capacity) and mechanical (e.g., elastic properties, thermal expansion, irradiation-induced dimensional changes, and irradiation-induced creep) material models for various graphite grades (e.g., H-451, IG-110, G-348, 2020, A3-3, and A3-27) are incorporated into Bison. Two benchmark problems are then exercised utilizing these new graphite-related capabilities: (1) modeling an Advanced Gas Reactor (AGR)-2 fuel compact, and (2) modeling the debonding of a particle-matrix interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Development of a Laser-Electron-Interaction Region for Compact High Repetition-Rate Mono-Energetic Laser Based Gamma Sources

Inverse-Compton scattering (ICS) is a well-established technique capable of producing monochromatic radiation over a large range of the electromagnetic spectrum. Most recently, ICS has attracted attention, owing to its compactness and possible coupling with novel acceleration methods to yield compact X- and 𝛄-ray sources. Under this award, we carried out a systematic optimization of the ICS and demonstrate via numerical simulation that combining ICS with a conventional superconducting linac could produce Watt-level average-power 𝛄 rays with high spectral brightness. The study is performed with idealized laser and electron-beam distributions and confirmed with the realistic distributions simulated from start-to-end numerical modeling of the electron accelerator. The studies led to the design and engineering design of an ICS interaction point for the FAST facility along with some beam measurement. Due to the cancellation of the program in Year 3 long with delay in the commissioning of the FAST facility, we were not able to demonstrate the generation of 𝛄 rays, we, however, procured some of the hardware, installed some of the required infrastructures and plan to eventually generate 𝛄 ray pending future interest on the topics and availability of funding.

43 PARTICLE ACCELERATORS↗

DOE OSTI2022

Irradiation of MiniFuel Targets Bearing TRISO Fuel Compacts (Status Report)

Irradiation testing of MiniFuel compacts bearing tristructural isotropic (TRISO) fuel particles was performed at Oak Ridge National Laboratory (ORNL) to support the development of Kairos Power’s (KP’s) fluoride salt–cooled high-temperature reactor concept. The fuel compacts were fabricated with TRISO fuel particles of different types—including low-enriched uranium oxide, uranium carbide (LEUCO), natural uranium oxide, uranium carbide (NUCO), and low-enriched uranium dioxide (LEUO 2 )—and inserted into MiniFuel irradiation targets. Five targets were assembled and inserted in the High Flux Isotope Reactor (HFIR) for four cycles. The data collected post-irradiation will provide experimental input to validate TRISO fuel performance models for high particle power operations. This report summarizes the completion of the HFIR irradiation, the as-irradiated numerical analysis, and the post-irradiation work performed to date. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 1-3, 4-3, 10-1, and 10-2

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support modeling improvement. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Determination of Average Burnup in AGR-3/4 Compacts 1-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite. These two goals were combined in the AGR-3/4 irradiation, which consisted of 12 independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. Two capsule types were used: a standard capsule and a so-called fuel body, in which the outer graphite ring included a floor and cap that fully enclosed the fuel. The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from the matrix and graphite at accident temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Compact, high-power superconducting radio-frequency accelerators for environmental applications

Electron-beam irradiation has been proven to destructively reduce or eliminate a wide variety of organic chemicals, viruses and bacteria from wastewater, as well as reducing sulfur and nitrous oxides emission from coal-fired power plants. It is estimated that there are approximately 30,000 such particle accelerators in use worldwide for industrial processes including surface and bulk processing of material, medical sterilization, and environmental remediation. Maximum beam power from accelerators used in these applications is currently limited to less than 500 kW, and a higher beam power is needed to reduce treatment costs. The market availability of continuous-wave (CW) electron-beam accelerators with power of the order of ~100 kW is also very limited. Superconducting radio-frequency (SRF) linear accelerators (linacs) are commonly used at basic research laboratories throughout the world due to their exceptionally high efficiency as compared to current industrial varieties. Recent advances in cryogenics, SRF thin films and high-power magnetrons allow for the design of increasingly compact and efficient CW SRF electron linacs in the energy range 1-10 MeV, with up to 1 MW of beam power. A new class of compact, high-efficiency electron-beam accelerators may provide cost-effective solutions for a range of industrial and environmental remediation applications, particularly with respect to tackling one such class of contaminants so-called forever chemicals, including per- and polyfluoroalkyl substances (PFASs), which are ubiquitous in a wide range of products and for which there is currently no effective destruction technology. We have recently demonstrated the key technologies required for such accelerators by: 1) operating a Nb3Sn SRF accelerating cavity cooled by commercial cryocoolers up to an accelerating gradient of 12.4 MV/m and 2) demonstrating the phase-locking as well as a high-efficiency power combining scheme for industrial magnetron transmitters. In this presentation, we will introduce the principles and benefits of MW-class SRF linacs, based on the conduction-cooled SRF technology that we have demonstrated for environmental remediation. We propose the development of a 4 MeV, 20 kW prototype to be built at Jefferson Lab as a first accelerator demonstrator unit. We will also present the results from samples study on the effect of electron-beam irradiation on so-called forever chemicals such as 1,4-dioxane and PFAS, using an existing multi-purpose 10 MeV, low-power, CW SRF linac at Jefferson Lab.

Ciovati, Gianluigi↗