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At least 415 records · Page 23

Concept Descriptions for the VTR Rabbit System and Driver Fuel Test Assemblies

Two of the experiment vehicles being developed for the Versatile Test Reactor (VTR) are presented here. The first is a rabbit system that will enable rapid insertion of small test capsules into the high fast flux of the VTR core for relatively short durations. The rabbit concept development includes the construction/demonstration of a near-full-scale system in a deep-water pool to demonstrate functionality, development of a concept of operations and initial procedures, and validation of thermal-hydraulic modeling. In addition, modeling efforts are underway to simulate the thermal and neutronic environment of a rabbit capsule. The second type of experiment vehicle presented here is a driver fuel test assembly for inserting fuel and materials tests into the core by replacing a driver fuel assembly. Here, a novel design for dismountable test assemblies is proposed for the VTR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Irradiation of Advanced Cladding Specimens in the High Flux Isotope Reactor: Capsule Designs and Test Matrix

The Advanced Fuels Campaign (AFC) has initiated the Advanced Reactor Cladding (ARC) irradiation campaign to generate irradiation performance data for candidate fuel cladding concepts. The campaign includes a diverse set of ferritic/martensitic steels, oxide dispersion strengthened (ODS) alloys, FeCrAlbased alloys, coated materials, and welded cladding specimens produced through multiple US Department of Energy (DOE) programs and international collaborations. Three complementary experimental thrusts comprise the campaign: tensile testing (ARC Tensile) to rapidly screen candidate alloys, fracture toughness testing (ARC Fracture) to evaluate irradiation effects on crack resistance, and tubular weld testing (ARC Weld) to quantify irradiation-induced changes in the mechanical performance of end cap welds. This report documents the irradiation campaign design, including the selected materials, specimen types, irradiation matrix, and capsule designs for irradiation within the High Flux Isotope Reactor (HFIR). A total of 14 irradiation capsules were developed to achieve target irradiation temperatures between 300°C and 600°C and doses up to 30 dpa. Thermal analyses were performed using finite element methods to establish capsule geometries capable of achieving the desired specimen temperatures while accommodating differences in specimen geometry and material properties. The resulting capsule designs provide the basis for irradiation of the AFC-ARC experimental matrix and subsequent post-irradiation examination to assess the effects of neutron irradiation on advanced cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Coupled Reactor and Engine Nuclear Thermal Propulsion Modeling Methodology

The design and development process of a Nuclear Thermal Propulsion (NTP) system requires extensive multiphysics modeling to couple the neutron physics and thermal feedback effects to determine the reactor’s power shape. Propulsion system performance codes utilize this power shape to determine NTP key performance parameters. While the power shape is heavily dependent on the temperature profile and geometry of the reactor, many analyses either assume a constant power shape, or use neutronics analysis to determine a power shape for a specific reactor configuration. The development of a coupling interface for a propulsion system performance code and a Monte Carlo neutron transport code (OpenMC) allows for the reactor power shape to be calculated in an iteration loop. The interface utilizes a file share system to transfer geometry dimensions, temperatures, and material identifiers to OpenMC, which is used to perform a neutron transport simulation of a design like the government Testing Reference Design reactor. The interface is then able to post-process the results from OpenMC and use the same file share system to share a power shape and other important neutron transport parameters to the system performance code. Initial results show that neglecting the changes to power shape when comparing reactor configurations can yield inaccurate results. Furthermore, utilizing propellants other than hydrogen gas can cause significant changes to the power shape, and thus, the thermal performance of a specific reactor design. This methodology is being expanded to allow for multiple families of NTP reactors to be analyzed, including block moderator, particle bed, and NERVA-derived reactors.

multiphysics coupling↗

Coupled Reactor and Engine Nuclear Thermal Propulsion Modeling Methodology

The design and development process of a Nuclear Thermal Propulsion (NTP) system requires extensive multiphysics modeling to couple the neutron physics and thermal feedback effects to determine the reactor’s power shape. Propulsion system performance codes utilize this power shape to determine NTP key performance parameters. While the power shape is heavily dependent on the temperature profile and geometry of the reactor, many analyses either assume a constant power shape, or use neutronics analysis to determine a power shape for a specific reactor configuration. The development of a coupling interface for a propulsion system performance code and a Monte Carlo neutron transport code (OpenMC) allows for the reactor power shape to be calculated in an iteration loop. The interface utilizes a file share system to transfer geometry dimensions, temperatures, and material identifiers to OpenMC, which is used to perform a neutron transport simulation of a design like the government Testing Reference Design reactor. The interface is then able to post-process the results from OpenMC and use the same file share system to share a power shape and other important neutron transport parameters to the system performance code. Initial results show that neglecting the changes to power shape when comparing reactor configurations can yield inaccurate results. Furthermore, utilizing propellants other than hydrogen gas can cause significant changes to the power shape, and thus, the thermal performance of a specific reactor design. This methodology is being expanded to allow for multiple families of NTP reactors to be analyzed, including block moderator, particle bed, and NERVA-derived reactors.

multiphysics coupling↗

Project Methods to Enhance Wellbore Cement Integrity with Microbially-Induced Calcite Precipitation (MICP) (Final Scientific/Technical Report)

The goal of this project was to develop improved methods for sealing compromised wellbore cement in leaking oil and gas wells, thereby reducing the risk of unwanted upward fluid migration. Novel methods for improving wellbore integrity, such as microbially induced calcite precipitation (MICP), can reduce leakage potential, improve the safety of fossil fuel extraction, improve the public perception of hydraulic fracturing, and promote environmentally-prudent unconventional oil and gas development. Microbes, with the urease enzyme, can catalyze the chemical reaction of urea hydrolysis to induce the precipitation of calcium carbonate which can be used as a cementitious material to seal leakage pathways. In this project, methods to promote robust bio-composite cementitious materials were designed and tested in the laboratory. Scale-up of those methods were tested in meso-scale reactor systems and in field applications. In this report, in Section One, we describe laboratory efforts to develop injection strategies to promote precipitation in wellbore analogs and determine the strength of the bio-composite cements as compared to fine cement. In Section Two, we describe the efforts to scale up the work and study the use of materials that can be used in field application, for example exploring the use of calcium chloride ice melt or urea fertilizer as source chemicals. In Section Three, the three field trials (methods and results) performed as part of the project are described and summarized. At the end of the report is a comprehensive summary and conclusion section which highlights the key findings of the project. The work performed during this project significantly advanced the technology readiness level (TRL) of the MICP wellbore sealing strategy.

02 PETROLEUM↗

Methods to Enhance Wellbore Cement Integrity with Microbially-Induced Calcite Precipitation (MICP) (Final Scientific/ Technical Report)

The goal of this project was to develop improved methods for sealing compromised wellbore cement in leaking oil and gas wells, thereby reducing the risk of unwanted upward fluid migration. Novel methods for improving wellbore integrity, such as microbially induced calcite precipitation (MICP), can reduce leakage potential, improve the safety of fossil fuel extraction, improve the public perception of hydraulic fracturing, and promote environmentally-prudent unconventional oil and gas development. Microbes, with the urease enzyme, can catalyze the chemical reaction of urea hydrolysis to induce the precipitation of calcium carbonate which can be used as a cementitious material to seal leakage pathways. In this project, methods to promote robust bio-composite cementitious materials were designed and tested in the laboratory. Scale-up of those methods were tested in meso-scale reactor systems and in field applications. In this report, in Section One, we describe laboratory efforts to develop injection strategies to promote precipitation in wellbore analogs and determine the strength of the bio-composite cements as compared to fine cement. In Section Two, we describe the efforts to scale up the work and study the use of materials that can be used in field application, for example exploring the use of calcium chloride ice melt or urea fertilizer as source chemicals. In Section Three, the three field trials (methods and results) performed as part of the project are described and summarized. At the end of the report is a comprehensive summary and conclusion section which highlights the key findings of the project. The work performed during this project significantly advanced the technology readiness level (TRL) of the MICP wellbore sealing strategy.

03 NATURAL GAS↗

Transformational Challenge Reactor preconceptual core design studies

In the nuclear industry, a manufacturing-informed design approach has the potential to yield the most benefit from advanced manufacturing. By leveraging advanced materials, data science, and rapid testing and deployment, manufacturing-informed design can drive down costs and development times, ultimately improving future commercial viability. This approach is being demonstrated in the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) program. Preconceptual design activities for TCR have been focused on analyzing and maturing four reactor core design concepts: two fast-spectrum and two thermal-spectrum systems. Furthermore, the designs were iteratively modified and analyzed, and subcomponents were manufactured in parallel over weeks instead of months or years. To meet key program initiatives (e.g., timeline and material use), several constraints—including fissile material availability, component availability, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small cores less than one cubic meter in volume with near-term viability. Additionally, the TCR program has made significant progress on development of advanced moderator materials such as yttrium hydride, advancing the feasibility of gas-cooled thermal spectrum systems using less than 250 kg of high-assay low enriched uranium (HALEU) and occupying less than 1 m3. Each of the two resulting thermal designs uses a different fuel form: traditional UO2 ceramic fuel and tristructural isotropic (advanced TRISO) fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems were assessed and summarized. Evaluation of the performance metrics for these two moderated designs has yielded the downselected TCR design: a TRISO-fueled and yttrium hydride moderated gas-cooled reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Report on Thermal Cycling Testing and Bimaterial Weld Development for a Passively Actuated Materials Surveillance Test Article

This report describes the continuation and conclusion of a thermal cycling test program designed to evaluate a family of passively actuated material surveillance test articles. Additionally, the report details a preliminary study of bimetallic welding between 316 stainless steel and TZM (tungsten-zirconium-molybdenum) alloy. The test program was ultimately successful in demonstrating the basic viability of the passively actuated test article concept as well as specifically validating the procedures developed to size specimens to mimic the mechanical response of operating components. The report discusses and summarizes the results of the complete test program which tested samples for more than 1,000 controlled thermal cycles. The weld development program included both stir friction and electron beam joining. The stir friction welding was ultimately successful, while the electron beam welding would require additional development. Ultimately, these test programs pave the way for implementing a materials surveillance program in future molten salt reactors (MSRs). The report conclusions outline a plan for transforming the research and developed described here into a definite set of practices and procedures for a surveillance program at a future operating reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Pu-238 Production Progress at Idaho National Laboratory From December 2022 to December 2023

Idaho National Laboratory (INL) has an ongoing effort to produce Pu-238 for NASA deep space missions. Recent work at INL has consisted of irradiation of Pu 238 production targets in the Advanced Test Reactor (ATR), generating more than an estimated 400 grams of Pu-238 heat source material between March and October 2023. Additionally, INL began qualifying Pu-238 production targets with a higher loading of Np-237 to further increase the production of Pu-238 production in later years. INL has also updated the analysis as a result of operational changes at ATR. One instance was updating the analysis to enable a single Pu-238 production target to be run in ATR’s South Flux Trap (SFT), rather than the previous seven production targets, to make use of a spare target from a discontinued design. Higher lobe powers in ATR were also analyzed due to potential changes in planned lobe powers.

07 ISOTOPE AND RADIATION SOURCES↗

Pu-238 Production Progress at INL From December 2022 to December 2023

Idaho National Laboratory (INL) has an ongoing effort to produce Pu-238 for NASA deep space missions. Recent work at INL has consisted of irradiation of Pu 238 production targets in the Advanced Test Reactor (ATR), generating more than an estimated 400 grams of Pu-238 heat source material between March and October 2023. Additionally, INL began qualifying Pu-238 production targets with a higher loading of Np-237 to further increase the production of Pu-238 production in later years. INL has also updated the analysis as a result of operational changes at ATR. One instance was updating the analysis to enable a single Pu-238 production target to be run in ATR’s South Flux Trap (SFT), rather than the previous seven production targets, to make use of a spare target from a discontinued design. Higher lobe powers in ATR were also analyzed due to potential changes in planned lobe powers.

07 ISOTOPE AND RADIATION SOURCES↗

Development of Conceptual Lead Cartridge Design to Perform Irradiation Experiments in VTR

This paper seeks to introduce the latest design of the Extended Length Test Assembly–Cartridge Lead (ELTA-CL) with associated thermal-hydraulic (TH) assessment and related experiment activities to support the critical component development performed by the ELTA-CL team (Los Alamos National Laboratory, Westinghouse Electric Company, and the University of New Mexico). The goal of the ELTA-CL program is to develop and validate an experimental capability to perform irradiation experiments in the Versatile Test Reactor (VTR) addressing Lead Fast Reactor (LFR) technology gaps, in support of the commercial development of advanced lead-cooled fast reactor concepts. Through a design maturation process and parametric study, a conceptual design is proposed to meet the requirements for material and corrosion testing. Thermal-hydraulic characteristics for the conceptual design at desired operating conditions are assessed with systems-level (one-dimensional) and computational fluid dynamics (three-dimensional) simulations. Along with the conceptual design work, experimental activities for the development of critical components such as the pump and flowmeter are undertaken. From both the modeling study and the experimental results, the design requirements of the Phase 1 ELTA-CL (e.g., 500°C and 2 m/s) are achievable with the current conceptual design. Additional design improvements and safety assessments at both steady-state and transient conditions for the final ELTA-CL design will be pursued.

42 ENGINEERING↗

Report on Hydrogen Content Measurements of Yttrium Hydrides

This report describes the hydrogen content measurements of yttrium samples as irradiated at Advanced Test Reactor as part of the Microreactor Program. Irradiated samples were prepared at the facilities of Analytical Research Laboratory (ARL) of Materials and Fuels Complex (MFC). Hydrogen content measurements were performed on reduced size specimens using an inert gas fusion analyzer. For a single sample, replicate samples were prepared and tested to improve statistics of results. Uncertainty analyses were conducted using two approaches to determine the hydrogen content variations in the samples. Hydrogen content of irradiated specimens were both lower and higher than the expected values, indicating stoichiometry variations. Hydrogen was detected in all samples, even for the cracked capsule’s specimens. Results also suggested a potential hydrogen redistribution is present between samples inside the capsules.

08 HYDROGEN↗

Testing and Characterization to Develop a Mechanistic Explanation for Unsaturated Drift of Fiber Optic Sensors during High-Dose Irradiation

The primary limitation for any optical fiber-based sensor for nuclear reactor applications is radiation-induced attenuation (RIA) of the transmitted and/or reflected signals. Based on several recent studies, RIA is tolerable for some fused silica optical fibers with the proper choice of sensing wavelength and fiber dopants. For extreme temperature applications (> 1000°C), sapphire optical fibers have been proposed; however, recent optical transmission measurements performed on bulk sapphire samples showed prohibitively large RIA. For some sensors, radiation-induced dimensional changes in the fiber materials can also cause significant drift. Moreover, the drift that was observed in numerous experiments performed in the High Flux Isotope Reactor (HFIR), the Advanced Test Reactor, the Massachusetts Institute of Technology Reactor, and other international facilities far exceeded what would be expected based on compaction of fused silica glass. Clearly, additional work is needed to better understand the origins of both RIA and radiation-induced drift in both silica and sapphire optical fiber-based sensors before these sensors can be reliably deployed for nuclear applications. This work evaluated the underlying mechanisms that may be responsible for RIA and drift in silica and sapphire materials. First, detailed characterization was performed on bulk fused silica glass samples that were previously irradiated to different neutron fluences at different temperatures to better understand the structural changes that drive radiation-induced drift in the absence of coating effects that are discussed later. Results show that the non-monotonic compaction that occurs with increasing neutron fluence continues up to fast neutron fluences approaching 10 22 n/cm 2 , which has important implications for physics-based models that may be used to compensate for the sensor drift. Initial Raman spectroscopy and synchrotron x-ray diffraction provide insights into the nature of the structural changes. Next, detailed characterizations were performed on silica fibers with various coatings that were subjected to several different thermal treatments. The hypothesis is that the coatings convert to carbon-rich materials that compact under irradiation, putting a large compressive strain on the fiber. Out-of-pile testing confirms that both polyimide and acrylate fiber coatings convert to glassy carbon (GC) materials when heated under inert conditions, and the degree of order (i.e., graphitization) increases with increasing temperature. The results provide increasingly strong evidence that the combination of polymeric coatings and inert (or vacuum) conditions render fiber optic sensors susceptible to significant radiation-induced drift that would not otherwise exist in uncoated fibers. Finally, transmission electron microscopy was performed on bulk sapphire samples that were irradiated to two neutron fluences at different temperatures to gain insights into the potential mechanisms driving the prohibitive RIA at higher neutron fluences and temperatures. Contrary to previous hypotheses, results show that scattering from radiation-induced voids cannot explain the observed RIA. Similarly, models for scattering losses from dislocation loops also do not agree with the experimental results. Instead, fitting to the experimental data shows that increased absorption from aluminum vacancy centers is the most likely explanation for the the prohibitively large RIA that was observed at high irradiation temperature and dose. In addition, the voids that formed in these single-crystal samples were found to align along the basal plane (a-axis) as opposed to that seen in previous observations of c-axis alignment in polycrystalline samples, which could have important implications for anisotropic swelling and other phenomena that could affect sensor performance at high neutron fluence.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Scaling for Nuclear Reactor System: Overview

A scaling analysis is essential for predicting the behavior of a reactor system under various operating conditions, as well as for ensuring safe and reliable experimentation and adequate dataset generation for supporting the system code model development and assessment in accordance with USNRC guidelines 1.203. Testing and modeling can be carried out at various scales, from small test units to full-scale reactors, to validate the predictions produced by the analysis. However, due to the cost and challenges associated with full-scale testing, the nuclear industry uses scaled test facilities—integral effect test (IET) and separate effect test (SET)—to identify potential design issues and optimize the design to ensure regulatory compliance for safer operation and anticipated accident management. In general, IET facilities are non-nuclear thermal hydraulics test facilities. In contrast, SET facilities can be non-nuclear testing (e.g., steam generator, containment, pump system SETs) and nuclear testing (mostly fuel-material irradiation testing). IETs are used to study the system-level thermal-hydraulic behavior of the reactor coolant system (RCS) under different anticipated operational occurrences (AOOs) and design basis accidents (DBAs) such as loss-of-coolant-accident (LOCA) and loss-of-power (i.e., station blackout) and loss-of-flow (i.e., failure of RCS pump). Several LOCA scenarios could happen in reactor DBAs, such as main steam line break (MSLB), SG tube rapture (SGTR), break in direct vessel injection (DVI), and other small breaks LOCA. Scaling analysis determines the appropriate size and operating conditions for the test facility. This involves analyzing the system's thermal-hydraulic behavior at different scales and identifying the key parameters that affect the thermal-hydraulic behavior of the reactor. These key parameters can be identified so that the scaled facility can replicate the expected conditions in a prototypical setting. The scaled facility can simulate the same flow rates, pressure gradients, heat transfer coefficients, and fluid properties as the prototype. Scaling analysis is used to ensure that the results from these tests can be scaled up to predict the behavior of a full-scale prototype with an acceptable range of scaling distortions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Experimental Methods for the Performance Test of Environment-Assisted Cracking (EAC) in Nuclear Industry and Examples through NSUF

Environmentally assisted cracking (EAC) is a critical failure mode in nuclear reactor environments, where materials are exposed to high temperatures, radiation, and corrosive media. Specific EAC studies in nuclear environments include stress corrosion cracking in primary water chemistry, irradiation-assisted stress corrosion cracking (IASCC) under neutron exposure, and hydrogen induced cracking (HIC) in reactor pressure vessels. The experimental methodologies of EAC include slow strain rate testing (SSRT), constant load, and cyclic loading techniques, conducted in autoclaves designed to simulate the reactor’s operational environments. This paper highlights several experimental case studies, especially on IASCC, conducted within the Nuclear Science User Facilities (NSUF).

36 - MATERIALS SCIENCE↗