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At least 343 records · Page 19

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DOE OSTI2022

Diagnostics for Burning Plasmas

Here, recent progress in the design and manufacturing development of in-vessel diagnostic components by the U.S. ITER team is reported. Design and manufacturing choices are driven by engineering challenges stemming from ITER’s burning plasma mission, which entails levels of plasma neutron emission, pulse lengths, and size scales well beyond the existing database. Challenges include port access limitations, neutron irradiation and activation risks, nuclear heating of components, complex interfaces, remote handling compatibility, and constraints on component and processing materials. Responses to these challenges are presented using examples from the low field side reflectometer, toroidal interferometer polarimeter, and electron cyclotron emission diagnostics. Relevance of the experience from ITER diagnostics engineering to the design of future integrated fusion facilities based on burning plasmas is addressed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

Advanced Test Reactor (ATR) fuel has been identified as a resource for high-assay low-enriched uranium (HALEU) production. A survey was performed on the published literature describing ATR fuel. The geometry of the fuel is complex; different parts of the fuel compact experience differing neutron flux and burnup. The literature is sparse, and access is controlled. Therefore, fundamental studies of fuel reprocessing must use a model fuel that represents the main chemical and structural features. Advanced chlorination, or chlorination with sulfur-chlorine bearing reagents is being investigated as way to separate the fuel from metal matrix alloys. A UAl x alloy will be fabricated with x = 3, 4, and 5. The potential chlorination of individual UAl x intermetallics will be assessed in the advanced chlorination process of Al-8001 and Al-6061 as well as a representative mixture. Initial studies will track the alloying elements of the Al, which are Si, Fe, Cu, Mn, Mg, Cr, Zn, and Ti, in addition to the U itself. Further studies will include fission product simulants. Because advanced chlorination solvents include sulfur, the chemistry of sulfur with major and minor constituents will also be investigated. The experimental work accompanied by neutronic calculations will allow the assessment of the feasibility of advanced chlorination to separate aluminum from uranium. If bench-scale testing appears promising, then small-scale tests in shielded facilities with irradiated cladding, lightly irradiated fuel, and spent nuclear fuel are recommended to track the complete inventory of fissile actinides, fission product impurities, and reagent solids and liquids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Feasibility Study of Advanced Manufacturing Techniques and Compositions of High Entropy Alloys

In line with the objectives of Department of Energy’s Office of Nuclear Energy Advanced Materials and Manufacturing technologies program (AMMT), this work focuses on new materials development and qualification research and development for next-generation, high-temperature nuclear reactors. While the currently qualified nuclear materials have demonstrated significant irradiation resistance at high temperatures, the community is still interested in materials that can sustain the harsh environments found in nuclear reactors for temperatures up to 1,000°C for nearly 100,000 hours. High entropy alloys (HEAs) have the potential to serve in these extreme environments of next generation nuclear reactors due to their unique phase transformation pathways and nanoscale and mesoscale microstructures. The current work focuses on understanding such nuclear energy relevant HEAs through a detailed literature survey and developing a decision matrix with criteria for identification of HEAs with may have most impact and value for further examination.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Steady state radiolysis–polyhalide accumulation products [Poster]

The study of steady-state radiolysis of polyhalides is crucial as it simulates the prolonged radiation exposure that materials (e.g., alloys, molten salts) experience in reactor environments. This research provides valuable insights into the long-term effects on material properties, chemical stability, and structural integrity. It is closely linked with fundamental radiation chemistry, which aids in predicting the chemical behavior of accumulation products. Understanding the initial speciation, yield, and chemical evolution of radiolysis products is essential for managing material degradation, controlling corrosion, and ensuring reactor safety.

36 - MATERIALS SCIENCE↗

DOE OSTI2023

MFC Transient Testing and Fuel Fabrication Capabilities

MFC Transient Testing Capabilities and Fuel Fabrication capabilities are summarized in this slide deck. Slides are included for TREAT and TREAT capabilities, EFF, FASB, FMF, ZPPR, AFF, and the Fuels Fabrication Modernization Capability. First use of this slide deck is for presentation at Westinghouse Electric Corporation, but there will be other uses later.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Nondestructive Evaluation of Concrete: Elastic Property Imaging Through Full Waveform Inversion

Concrete is a major construction material worldwide and plays a crucial role in the nuclear industry. The elastic properties of concrete are prone to change and degrade while in service, as it is often subjected to extreme operational and environmental conditions. An accurate evaluation of concrete's elastic properties is thus essential to ensure structural integrity and safety. This is especially true for concrete in nuclear power plants, where irradiation effects significantly impact concrete mechanical properties. There are various methods to assess these properties, with ultrasound-based techniques showing high potential due to their nondestructive nature, cost-effectiveness, and safety. While several nondestructive evaluation methods exist, most rely on idealizations such as assuming homogeneous material and plane wavefronts. In this work, we address these issues by introducing an ultrasound-based nondestructive method aimed at reconstructing spatially varying images of concrete mechanical properties. By accurately modeling wave physics, including scattering and reflection, we overcome several of the aforementioned idealizations and aim to utilize the full waveform for imaging material properties through depth, resulting in more reliable images. Full waveform inversion (FWI) was first introduced by geophysicists to reconstruct subsurface elastic property images. The goal is to minimize the difference between simulated and recorded wavefield signals, often through gradient-based optimization algorithms. While FWI is primarily conducted using the acoustic approximation of the wave equation, few works focus on elastic FWI, where the goal is to reconstruct images of not only the pressure wave speed but also the shear wave speed and density (or their equivalents). This work explores the potential of using elastic FWI to predict concrete mechanical properties as an initial effort for a more accurate monitoring of concrete conditions in service. Reconstructing images of different elastic parameters enables more specificity and accurate condition diagnosis. This paper will detail this approach and provide examples demonstrating the effectiveness of elastic FWI in reconstructing comprehensive maps of concrete mechanical properties.

42 - ENGINEERING↗

DOE OSTI2025

Characterization of Pinhole Collimators for High-Resolution Gamma Imaging of Irradiated Fuel

Post-irradiation examination (PIE) of nuclear fuels requires imaging tools capable of resolving isotopic and spatial features with high throughput. This project contributes to a proof-of-concept effort aimed at advancing gamma emission tomography (GET) by evaluating novel fine-aperture pinhole collimators. Two Rose’s metal collimators, 100 µm (20° acceptance angle) and 350 µm (30° acceptance angle), were prototyped and characterized for their effectiveness in transporting gamma rays through the pinhole aperture. To support data collection, a Python-based data acquisition system was developed to coordinate a rotation stage, linear stage, and CZT detector, reducing latency in high-rate gamma event logging to one second per acquisition. Queue-based file writing enabled seamless real-time data capture for count rates up to 35,000 counts per second (cps). List-mode parsing algorithms were implemented to differentiate single and simultaneous gamma interactions for future tomographic reconstruction. Detector response was evaluated in both spectroscopy and list mode acquisition methods across varying source distances to confirm absolute and collimator efficiencies. Preliminary efficiency figures suggest effective collimation of gamma-rays with energies below 700 keV, with ~4% residual intensity through the aperture for Cs-137. The impact of collimator geometry on image quality is currently being evaluated. This groundwork supports the ongoing development of a sub mm resolution cone-beam CT system for imaging fuel phantoms, an essential step toward improving GET efficiency and accelerating nuclear fuel qualification efforts.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2021

Testing of an Ultrasonic Thermometer Developed at the Idaho National Laboratory

Ultrasonic thermometry has great potential to improve temperature measurements in irradiation experiments performed in nuclear test reactors. The flexibility of the measurement makes it applicable to a wide range of potential test conditions, in terms of temperature, radiation, and chemical environment. However, the range of several typical ultrasonic thermometer configurations are not well defined. The research described in this document attempts to define an operational envelop for multiple configurations and sensor materials.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2021

Probing Microstructure-induced Swelling & Thermal Property Changes in Ion-Irradiated Oxide Fuels using Laser-generated Surface Acoustic Waves

Radiation of fuel rods with neutrons during a fission reaction in a nuclear reactor is known to affect the material microstructure of the fuel as a result of fission fragment damage, and lead to the formation of atomic-level defects such as voids, dislocation loops, and lattice swelling from fission-gas release. A consequence of the radiation-induced damage in nuclear fuels is the drastic alteration of material properties, in particular, the reduction of the thermal conductivity – a key parameter that governs the transport of thermal energy released from fissile fuel atoms to the surrounding coolant. Swelling can induce high stresses and ultimate failure of the cladding. A fundamental understanding of the role of radiation-induced damage on the thermal transport and mechanical properties of nuclear fuels is therefore critical for efficient, reliable and safe operation of a nuclear power plant. As opposed to conventional post-irradiation examination techniques that are often time-consuming and require significant sample preparation, laser-based characterization methods have emerged as promising non-contact, non-destructive tools to measure the evolution of microstructure-induced material property changes. Moreover, laser methods offer the promise of in-situ characterization while the material is being irradiated. This project aims to develop an improved understanding of the impact of radiation-induced changes to material microstructure on the thermal and elastic properties of nuclear materials. A laser ultrasonic technique, known as the transient grating (TG) spectroscopy method, is used to simultaneously measure thermal diffusivity and elastic properties in unirradiated and ion-irradiated oxide nuclear fuel samples. The results will complement on-going investigations on electron- and phonon-mediated thermal transport in nuclear materials and will provide foundational work for incorporating the influence of defects in fuel performance codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Comparing structure-property evolution for PM-HIP and forged alloy 625 irradiated with neutrons to 1 dpa

The nuclear power industry has growing interest in qualifying powder metallurgy with hot isostatic pressing (PM-HIP) to replace traditional alloy fabrication methods for reactor structural components. But there is little known about the response of PM-HIP alloys to reactor conditions. This study directly compares the response of PM-HIP to forged Ni-base Alloy 625 under neutron irradiation doses ~0.5–1 displacements per atom (dpa) at temperatures ranging ~321–385 °C. Post-irradiation examination involves microstructure characterization, ASTM E8 uniaxial tensile testing, and fractography. Up through 1 dpa, PM-HIP Alloy 625 appears more resistant to irradiation-induced cavity nucleation than its forged counterpart, and consequently experiences significantly less hardening. This observed difference in performance can be explained by the higher initial dislocation density of the forged material, which represents an interstitial-biased sink that leaves a vacancy supersaturation to nucleate cavities. These findings show promise for qualification of PM-HIP Alloy 625 for nuclear applications, although higher dose studies are needed to assess the steady-state irradiated microstructure.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Fundamental Phenomena and Applications of Swift Heavy Ion Irradiations

This review concentrates on the specific properties and characteristics of damage structures generated with high-energy ions in the electronic energy loss regime. Irradiation experiments with so-called swift heavy ions (SHIs) find applications in many different fields, with examples presented in ion-track nanotechnology, radiation hardness analysis of functional materials, and laboratory tests of cosmic radiation. The basics of the SHI-solid interaction are described with special attention to processes in the electronic subsystem. The broad spectrum of damage phenomena is exemplified for various materials and material classes, along with a description of typical characterization techniques. This review also presents state-of-the-art modeling efforts that try to account for the complexity of the coupled processes of the electronic and atomic subsystems. Lastly, a brief discussion at the end of this review on SHI phenomena induced by fission fragments in nuclear materials will highlight the relevance of energetic ion irradiation in nuclear technology applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Status on Development of Graphite Analytical Tool (GAT)

The DOE-ART Graphite R&D program has been generating significant amounts of irradiated and unirradiated graphite data since 2006 when the program was part of the DOE NGNP (Next Generation Nuclear Plant) Project. This data includes critical irradiation creep and irradiated material property changes from the Advanced Graphite Creep (AGR) experiment as well as significant amounts of data on unirradiated material property values on several current nuclear graphite grades (Baseline program). Previously, Idaho National Laboratory has developed an internal analysis tool to assist with analysis of the unirradiated and irradiated data. The Graphite Analytical Tool (GAT) is intended to provide easy access to the graphite data in the form of comparing unirradiated and irradiated material property changes, comparison of material property differences between various nuclear graphite grades, and illustrate trends within the irradiated and unirradiated data generated within the DOE-ART Graphite R&D program. This report summarizes the progress to-date on the development of this analytical tool.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

In situ observation of irradiation-induced enhancement to the desorption pressure of zirconium hydride in a nuclear reactor

We quantify the effect of a nuclear-reactor environment on the hydrogen isotope equilibrium vapor pressure over pure zirconium and zirconium hydride. A vacuum-sealed capsule containing a zirconium foil with 6 atom% deuterium was irradiated at a neutron flux of ~10 14 cm -2 s -1 at the University of Missouri Research Reactor (MURR). The internal stainless-steel (SS) sample holder acted as the heat source via gamma absorption. To measure low desorption pressures in a high-flux environment, we developed a method to transduce pressure from the measured sample temperature during irradiation, calibrating with known deuterium pressures in unirradiated capsules at various heating powers using an internal filament-heated system designed to mimic irradiation-induced heating. Our temperature-pressure transduction method operates similarly to a Pirani or thermocouple pressure gauge. The in-reactor measurements revealed a roughly 4-fold enhancement in desorption pressure after only 6 h of irradiation (~2 × 10 18 cm -2 neutron fluence) compared to thermal desorption in control experiments, indicating a nonthermal contribution from neutron irradiation. The slower temperature/pressure stabilization rate in the reactor suggests that desorption pressure enhancement increases with neutron fluence. Further, this enhancement signifies increased solubility of hydrogen isotopes in zirconium during irradiation. We propose that high-energy neutron collisions with hydrogen isotopes in hydrides lead to their decomposition at lower temperatures, supersaturating the surrounding αZr lattice and resulting in higher desorption pressure, which continues to rise as more hydrides dissolve with increasing neutron fluence.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Completion and Testing of a New High-Temperature Cell Holder to Support High Dose Rate Electron Beam Irradiations of Select Sulfur Chlorides (S2Cl2 and SOCl2) for Advanced Low Temperature Chlorination of Zirconium-based Used Nuclear Fuel Cladding

Short communication on the completion and testing of a new high-temperature cell holder to support high dose rate electron beam irradiations of select sulfur chlorides (S2Cl2 and SOCl2) for advanced low temperature chlorination of zirconium-based used nuclear fuel cladding.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the ATR

The Advanced Test Reactor (ATR) is a light water reactor with aluminum-clad driver fuel. A primary mission of the ATR is to support the next generation of nuclear reactors. This support necessarily requires irradiation of advanced materials such as sodium, fuel salts, and metal eutectics. Irradiation of advanced materials in the ATR environment presents a challenge when completing accident analyses and demonstrating compliance to the Safety Analysis Report (SAR). Many advanced materials have the possibility to react with the ATR protective barriers such as the cladding or primary coolant system (PCS) boundary during postulated accident scenarios. Further, molten fuel experiments fall outside of the standard regulatory framework for dose consequence analyses. ATR is currently developing new safety analysis methods to support irradiation of advanced materials. The primary considerations for this development are 1) experiment containment design requirements, 2) primary coolant system response to an experiment containment failure, and 3) dose analyses for molten fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Report on Preliminary Detailed Experimental Plan for Neutron Irradiation of A709 at ATR and HFIR

Advanced nuclear power technologies will use higher temperatures as a means to extract energy at a higher efficiency than current plants and therefore put a larger demand on the structural materials. Improved performance of structural materials could enable greater safety margins, longer plant lifetimes, and reduce maintenance costs. Alloy 709 is championed as the next generation of austenitic alloys for advanced nuclear reactors. In parallel to the ASME code case pursued, the AMMT program is initiating a neutron irradiation campaign to provide first-of-a-kind engineering data to establish operational design parameters and how the mechanical response is modified by environmental factors. This document refines the AMMT neutron irradiation campaign to a 4 year program to support the generation of creep knockdown factors for Alloy 709 and welded Alloy 709. The campaign is divided among two national laboratories, Oak Ridge National Laboratory and Idaho National Laboratory, to use the strengths of each laboratory. Through a cooperative plan, time-independent properties and time-dependent properties will be obtained across a large temperature window, nominally 300°C to 800°C, damage levels up to 10 dpa, and with and without the impacts of transmutation produced helium.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Forced flow transient safety analysis of irradiation device with adjustable orifice for research reactor fuel assemblies

The Belgium Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) has several irradiation devices or rigs that are dedicated to the fuel performance and qualification demonstration testing of research reactor fuels. In support of the U.S. High Performance Research Reactor (USHPRR) LEU conversion project, a new flexible irradiation apparatus, MUSTANG-R, has been constructed. SCK CEN has completed the design and safety study, in cooperation with Idaho National Laboratory (INL) and Argonne National Laboratory (ANL), to allow for the irradiation testing of a full-size fuel assembly in a 200 mm diameter channel in the BR2 reactor. The moveable valve is a key design feature of the device and acts like an adjustable orifice enhancing or restricting the flow through a coolant channel inlet located in the BR2 upper plenum. This moveable valve allows the flow through the device to be adjusted prior to each BR2 cycle to obtain the necessary conditions for the fuel qualification test. This ensures accurate and representative thermal-hydraulic conditions of the fuel design are achieved. The device was designed and qualified as passively safe, implying verification by a combination of mechanical and thermal-hydraulic analysis and testing. This includes characterization of the safety margin required for a scenario where the moveable valve is assumed to be erroneously closed during irradiation. A simplified and conservative method is proposed for analyzing the corresponding forced flow transient using a critical heat flux criterion. In conclusion, this allows the required minimum valve opening to be determined for the experiments' design and safety studies.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the Advanced Test Reactor

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗