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DOE OSTI2023

Electropotential Verification for Nuclear Safeguards

The international safeguards regime desires methods to efficiently verify that facilities are only performing declared activities. Electropotential verification (EPV) is a newly proposed technique that was tested for its feasibility to perform facility design information verification (DIV) and verification of spent nuclear fuel while in a cooling pool. EPV works by passing a constant, low voltage current through a conductive system (facility infrastructure of nuclear fuel assembly) and measuring the resulting voltage at various places throughout the infrastructure in order to establish a baseline. Changes made to the system affect these voltage readings, which will deviate from the baseline and indicate that a change to the system was made. For facility DIV, it appears feasible that changes in configuration of the system’s grounding can be detected in real-time, and the location of the change can be inferred from the measured intensity of the change in voltage. Determination of whether or not spent fuel was present in a fuel rod, as well as the presence/absence of a fuel rod from an assembly using EPV, proved unsuccessful with the sensitivity of instrumentation used in this study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY23) Appendix F: Cyclic Integrated Reversible-Bending Fatigue Tests

Spent nuclear fuel (SNF) assemblies must be shipped to other sites for processing and disposal. During shipment, the fuel is typically oriented horizontally, and the fuel rods are subject to periodic alternating loads related to the movement of the vehicle that results in alternating bending of the SNF fuel rods. The number of bending cycles is related to the length of the shipping route, with longer routes producing more cycles. Because it is well-known that cyclic loads can produce failures, even when the stress and strain imposed are below the material’s yield point, investigation of the SNF’s fatigue behavior is prudent. This report discusses the results of fatigue testing conducted at Oak Ridge National Laboratory (ORNL) using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) created by the US Department of Energy Office of Nuclear Energy for the High Burnup Spent Fuel Data Project and its sister rods. The results of the tests are compared with results obtained over the last decade using the same CIRFT for the US Nuclear Regulatory Commission (NRC).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Completion of Critical Experiments with Molybdenum Sleeves at Sandia

Sandia National Laboratories (SNL) and the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) have collaborated on the design and execution of a set of critical experiments that explore the effects of molybdenum in water moderated fuel-rod arrays. The molybdenum is included as sleeves (tubes) on some of the fuel rods in the arrays. The fuel used in the experiments is known at Sandia as the Seven Percent Critical Experiment (7uPCX) fuel. This fuel has been used is several published benchmark evaluations in including LEU-COMP-THERM-78 and LEU-COMP THERM-080.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Panel Session 145: Challenges and Opportunities in Establishing a System for Transportation of SNF/UNF

Issues associated with establishing a SNF Transportation System in the U.S. are complex. In part, this is because in the US when discussing SNF, there are two distinct categories. The two categories represent significant differences in ownership, transportation-regulator, and even the physical characteristics of the SNF. Specifically, the physical differences in the SNF include size of the rods, associated quantities to be shipped, cladding material, and radionuclide content/enrichment. The first is commercial SNF (i.e., that associated with nuclear power plants), owned by utilities, until a permanent repository is built with DOE obligated by law to then assume the responsibility for shipping and ownership of the SNF. Since no-permanent repository currently exists, most commercial SNF is stored near the reactor in which it was installed, hence requiring no shipments. Although current shipments of commercial SNF are very limited, the few that do occur are generally performed to support a utilities SNF consolidation effort utility and for post irradiation examination/testing. However, with approximately 83,000 metric tons currently being stored at more than 70 sites around the U.S., when a permanent repository becomes available, the required number of shipments will be extensive. The second category is DOE and research reactor fuel. These fuel rods are generally much smaller in length than commercial fuel rods, with the quantity of rods transported in a shipment significantly less than the number of rods in a typical commercial fuel assembly. Currently, most of the SNF shipments within the U.S come from this category. DOE is currently managing about 2500 metric tons of heavy metal of SNF. This panel focused on updates and status of transportation of SNF/UNF. Panelists provided updates and information on continuing impacts, risk assessment, equipment (rail cars, handling, securement), possible rail routing, and proposed inspection equipment/routing prior to shipments. Panelists with presentations: Preparing for Nuclear Waste Transportation (Daniel Ogg); Challenges and Opportunities in Establishing a System for Transportation of SNF/UNF (Kathy Langan); UNF Transportation: Challenges and Opportunities (Michael Valenzano)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Amphiphilic Block Copolymers for Flocculation and Hydrophobization of Legacy Waste Suspensions in Flotation Driven Dewatering Operations

There exists a degrading legacy fuel pond at the Sellafield site (UK), known as the First-Generation Magnox Storage Pond (FGMSP). After being placed into a passive care and maintenance regime resulting in a long storage period, the fuel rods, primarily their cladding, corroded in the pond forming magnesium hydroxide based sludges and suspended material. These ponds are a grave concern to the British government in terms of hazard and risk reduction. Therefore, decommissioning of these ponds is a top priority beginning with the contaminant retrievals process [1]. The pond has accumulated significant quantities of waste materials amongst the skips of fuel, including but not limited to: large inventories of corroded Magnox sludge, fuel rod fragments, metal fragments (from fuel skips), concrete degradation products (from the pond infrastructure), wind-blown sand, and other materials such as bird guano and animal remains. This challenge requires a chemically robust technology to complete sludge retrievals [2]. As nuclear is different, a more stringent process operation criteria is required. From the criteria, flotation was selected as a viable technology. Flotation involves the application of collector molecules which modify the hydrophobicity of suspended particles allowing them to adsorb to rising bubbles. As some particles lack inertia for flotation due to their size, dual flocculant/collector agents can be deployed for greater particle recovery, in this case, amphiphilic diblock copolymers. Two copolymers of Poly(acrylic acid)-b-poly(n-butyl acrylate) (or PAA-b-PnBA), of different hydrophobic chain lengths (PBA) were synthesized for flotation campaigns. Flotation has shown promise to be a valuable rapid dewatering strategy for decommissioning of legacy waste ponds whilst upholding the required operational criteria. This research has shown: PAA-b-PnBA copolymers promote flocculation of Mg(OH){sub 2} particulates. Diblock copolymers retain more fluid than traditional surfactant based collectors wrt. relative particulate recovery performance- where longer PnBA chain length polymers performed best. Performance appears to be hydrodynamically limited due to over-flocculation of Mg(OH){sub 2}. Potential to combine with sedimentation for high particle recovery with low fluid carryover.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Wireless Instrumented RB Experiment Preliminary Design and Analysis

The ability to deploy new nuclear fuels for current or future reactor concepts requires carefully designed experiments to generate data to support fuel qualification. Ideally these experiments would include state of-the-art sensing to maximize the amount of in situ data that can be collected during operation. Furthermore, advanced reactor systems can take advantage of integrated in-core sensing technologies to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator’s understanding of limiting peaking factors. Before any novel sensing technologies can be readily adopted for nuclear applications, they must first demonstrate acceptable performance in test reactors. This report summarizes the preliminary design and analysis of the most highly instrumented irradiation experiment ever performed in the removable beryllium (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Wireless Instrumented RB Experiment 2021 (WIRE-21) will test a wide range of sensors including wireless sensors being developed by Westinghouse Electric Company (WEC) that could provide in situ measurements of peak fuel temperatures and fuel rod pressurization due to fission gas release. The ability to wirelessly transmit a signal through the fuel rod’s cladding is critical to improving fuel monitoring capabilities without requiring signal penetrations through the cladding pressure boundary, which would significantly impact fuel fabrication, handling, and operation. Other sensors that will be tested in WIRE-21 include an array of thermocouples, self-powered neutron detectors (SPNDs), and spatially distributed fiber-optic temperature sensors. More generally, WIRE-21 will establish a flexible irradiation vehicle design to allow accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. This report summarizes the mechanical design for WIRE-21, the experimental test matrix, initial neutronic and thermal design analyses, and the active monitoring and control system enhancements necessary to support testing of advanced sensor technologies. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) relevant to light water reactors (LWRs), but the flexible design of the experiment vehicle allows much higher operating temperatures (>1,100°C). Neutronic calculations determine the neutron flux conditions as well as the nuclear heating within the experiments. These results are used as inputs to detailed thermal finite element calculations, which are required to evaluate the complex, three-dimensional heat transfer that occurs within WEC’s wireless sensor enclosures. Initial results show that the temperatures of the sensors’ enclosures and the metal bellows can be operated near the temperature range of LWR coolants and cladding while simultaneously increasing the temperature of a surrogate fuel material to values in the range of 800–1200°C to simulate centerline fuel temperatures during LWR operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

REGAL International Program: Analysis of experimental data for depletion code validation

The Rod-Extremity and Gadolinia AnaLysis (REGAL) Program is a joint international effort to expand the nuclide inventory experimental data for irradiated nuclear fuel, with a specific focus on addressing two challenging needs associated with the characterization of modern, high duty, nuclear fuel. The first challenge is filling the gaps in experimental nuclide inventory data for gadolinia (UO 2 –Gd 2 O 3 ) fuel rods. The huge absorption cross sections of Gd-155 and Gd-157 in the Gd dopant in these rods lead to atypical spatial self-shielding patterns and have an impact on the neutronic environment within the fuel assembly compared to regular UO 2 fuel rods. Additionally, the second challenge is investigating the impact of burnup gradients at rod extremities on fuel composition and neutron leakage, to provide relevant experimental data for assessing computational capabilities to model such impact. A benchmark has been defined as a first step in the development of best-estimate models in the preliminary phase of the experimental data evaluation. Comparison of experimental results obtained in Phase I of the program for two measured pressurized water reactor (PWR) samples, one UO 2 and one UO 2 –Gd 2 O 3 sample, with calculated results obtained with different computational tools based on the defined benchmark are presented and discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Metallographic examinations and hydrogen measurements of high-burnup spent nuclear fuel cladding

In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). Here, t o investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

BISON: A Flexible Code for Advanced Simulation of the Performance of Multiple Nuclear Fuel Forms [Slides]

As fuel vendors and designers pursue the development of advanced reactors or the increase in burnup limits for existing reactors, advanced computational tools are necessary to understand the fuel performance. BISON, a fuel performance code developed primarily at Idaho National Laboratory, which is built upon the Multiphysics Object-Oriented Simulation Environment (MOOSE), provides capabilities to analyze multiple nuclear fuel forms in a wide variety of dimensions. Since its inception, BISON has been used to investigate the performance of light-water reactor fuel rods, accident tolerant fuel concepts, metallic and mixed-oxide (MOX) fuels for fast reactors, plate fuels for research reactors, and tri-structural isotropic (TRISO) fuel particles. This talk will provide a history of BISON, highlights of major development milestones, tributes to key contributors, and applications of its use to various fuel forms in one-, two-, and three-dimensions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Defining Probe Closure Offset and Scan Stitching Methodology for Element Contact Profilometry in the Hot Fuel Examination Facility

The Element Contact Profilometer (ECP) is a non-destructive post-irradiation examination (PIE) method established in the Materials and Fuels Complex (MFC) Hot Fuel Examination Facility (HFEF). The ECP suspends a fuel rod with a standard end-fitting in a vertical orientation and draws it between two opposed Sony probes with sapphire rod tips, which measure the diameter directly as a function of vertical stage position through contact profilometry. The ECP is equipped with concave rollers, which pinch the rod to maintain a consist position between these sapphire probe tips. A schematic of the analysis head is shown in Figure 1. The gripper that suspends the rod end-fitting is able to rotate to allow measuring of any azimuthal angle around the rod. This system is used to characterize the diameter of full-length light water reactor (LWR) irradiated fuel rods at multiple azimuthal angles in support of the Advanced Fuels Campaign (AFC).

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Magnification Determination for AGHCF Metallographic Images

Metallographic images generated in the Alpha Gamma Hot Cell Facility (AGHCF) have been used for quantitative and qualitative characterization of irradiated metallic fuel pins. Individual high-magnification images (micrographs) contained in AGHCF file folders and experimental notebooks are in the form of polaroid photographs labeled with AGHCF identification number, magnification, and surface condition (as-polished or etched). Low magnification composite images (photomosaics) are also available for fuel cross sections, for radial strips from the fuel center to the cladding outer diameter, and for circumferential strips of the cladding and outer fuel region. The individual images do not have scale bars. Only a few of the composite images contain scale bars. The purpose of this work is to describe and apply methods for checking the magnification (50X to 500X) of individual images and for determining the magnification (~25X to ~100X) of composite images. The metallographic images used to verify and/or determine magnification were generated during the 1989 to 1993 timeframe for cross sections of U-10Zr/HT9 fuel rods irradiated in EBR-II and for cross sections of irradiated U-10Zr/HT9 fuel-rod samples following elevated-temperature tests conducted in the out-of-pile AGHCF Fuel Behavior Test Apparatus (FBTA). Procedures are documented in the AGHCF Operations Manual for preparation of metallographic samples, for calibration verification, and for constructing composites from individual images. Calibration verification of magnification was performed at least semiannually. One standard used for calibration verification was a glass slide with vertical lines indicating distances of 1 mm, 0.1 mm and 0.01mm. Guidance is provided in the current work for determining magnification from images of the standard. In addition to the periodic calibration verification, consistency checks are recommended to verify the magnification of images taken between the routine calibration verifications. For cladding samples with or without significant fuel-cladding chemical interaction (FCCI), images taken at different magnifications (e.g., 150X and 200X) of the unaffected cladding thickness can be compared. This process is independent of changes in cladding thickness due to swelling and creep. For images taken at only one magnification of cladding regions with no FCCI, the cladding thickness determined from the images may be compared to the nominal as-built cladding thickness (15.0 ± 0.5 mils for examples used in current work)). This procedure works best for cladding that has not experienced significant swelling and/or creep (e.g., ≤2% circumferential strain at the cladding outer surface). Composite images were constructed by pasting together higher magnification images and photographing the composite to generate a negative from which the hard-copy photograph was developed. The primary purpose of the composite images was to identify interesting areas of the fuel and cladding for imaging at higher magnification. However, procedures are recommended in the current work for determining composite magnifications to allow quantitative characterization of the fuel. Composites are not recommended for determining cladding thickness or FCCI depth because magnifications are too low (generally ≤110X) and image contrast is low. Fuel cross-section composites were created by pasting together 50X or 75X images. The nominal magnification for the photographed fuel cross-section composites is 25.4X. A more precise magnification can be determined by comparing the average cladding outer diameter measured from the composite to the average cladding outer diameter measured by profilometry. Radial-strip composites were constructed by pasting together high-magnification (150X or 250X) images. The nominal magnification of radial-strip composites photographs is about 100X. A more precise method for determining the magnification of radial strips is to compare the cladding thick

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Quantitative imaging and automated fuel pin identification for passive gamma emission tomography

Compliance of member States to the Treaty on the Non-Proliferation of Nuclear Weapons is monitored through nuclear safeguards. The Passive Gamma Emission Tomography (PGET) system is a novel instrument developed within the framework of the International Atomic Energy Agency (IAEA) project JNT 1510, which included the European Commission, Finland, Hungary and Sweden. The PGET is used for the verification of spent nuclear fuel stored in water pools. Advanced image reconstruction techniques are crucial for obtaining high-quality cross-sectional images of the spent-fuel bundle to allow inspectors of the IAEA to monitor nuclear material and promptly identify its diversion. In this work, we have developed a software suite to accurately reconstruct the spent-fuel cross sectional image, automatically identify present fuel rods, and estimate their activity. Unique image reconstruction challenges are posed by the measurement of spent fuel, due to its high activity and the self-attenuation. While the former is mitigated by detector physical collimation, we implemented a linear forward model to model the detector responses to the fuel rods inside the PGET, to account for the latter. The image reconstruction is performed by solving a regularized linear inverse problem using the fast-iterative shrinkage-thresholding algorithm. We have also implemented the traditional filtered back projection (FBP) method based on the inverse Radon transform for comparison and applied both methods to reconstruct images of simulated mockup fuel assemblies. Higher image resolution and fewer reconstruction artifacts were obtained with the inverse-problem approach, with the mean-square-error reduced by 50%, and the structural-similarity improved by 200%. We then used a convolutional neural network (CNN) to automatically identify the bundle type and extract the pin locations from the images; the estimated activity levels finally being compared with the ground truth. The proposed computational methods accurately estimated the activity levels of the present pins, with an associated uncertainty of approximately 5%.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Results of the CHF-SERTTA In-Pile Transient Boiling Experiments at TREAT

To advance the current understanding of heat transfer between nuclear fuel rod cladding and the surrounding coolant under transient scenarios involving critical heat flux (CHF), the CHF-Static Environment Rodlet Transient Test Apparatus (CHFSERTTA), an in-pile experiment, was developed for and irradiated in the Transient Reactor Test (TREAT) facility at Idaho National Laboratory. The CHF-SERTTA experiment consists of a hollow, borated stainless steel rod submerged in a static water pool. The rod, heated via the (n, a) reaction in boron-10, was designed for investigating in-pile transient heat transfer between the rod surface and the coolant, avoiding the complexity of including the heat transfer between the fuel, gas gap, and cladding in typical light-water reactor (LWR) fuel rods. The instrumentation package for CHF-SERTTA focused on capturing those data that best describe the heat transfer behaviour between the rod and coolant. This paper outlines the CHFSERTTA experiment design and presents the initial results of the experiment campaign.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

WIRE-21 Sensor Irradiation Experiment Ready for HFIR Insertion

The ability to deploy new nuclear fuels for current or future reactor concepts requires a wealth of data regarding fuel performance during normal operation, anticipated operational occurrences, and design-basis accidents. Most of these data have historically been collected during experiments in materials test reactors, ideally with online instrumentation to collect as much data as possible. However, advanced instrumentation could also allow for in situ monitoring of fuel operating conditions during commercial reactor operation to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator understanding of limiting peaking factors. The latter approach would complicate fuel handling, particularly during refueling, unless the instrumentation could be placed inside the fuel rods and transmitted wirelessly to a receiver located outside the fuel’s primary pressure boundary. To this end, Westinghouse Electric Company (WEC) developed wireless sensors based on inductive coupling that can transmit information regarding fuel centerline temperatures and rod internal pressures wirelessly from within a fuel rod to a nearby instrument thimble. After testing these sensors in lower-power university research reactors, the next step is to perform high neutron fluence testing to characterize the performance of these wireless sensors under conditions that are more representative of the intended application—in this case, light-water reactors (LWRs). The removable Be (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) provide the neutron flux, experiment volume, and access to instrument leads required to achieve these sensor testing goals. This report summarizes the design, analysis, and assembly of the Wireless Instrumented RB Experiment 2021 (WIRE-21). This is the most highly instrumented irradiation experiment ever performed in HFIR. The experiment will use seven different sensing techniques to measure temperature, pressure, neutron flux, and neutron fluence during reactor operation. In addition to WEC’s wireless temperature and pressure sensors, WIRE-21 includes an array of thermocouples, self-powered neutron detectors, spatially distributed fiber optic temperature sensors, passive SiC temperature monitors, and flux wires. The design of WIRE-21 and the cabling that was installed in HFIR also provide the infrastructure to enable accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting component temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) that would be expected in the plenum region of LWR fuels, except for the active sensing region of the wireless temperature sensor, which is targeting LWR fuel centerline temperatures (~800–1,100°C). WIRE-21 was successfully assembled, passed all nondestructive examination, and was delivered to HFIR for insertion during upcoming cycle 498 (April 2022).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

BISON validation to in situ cladding burst test and high-burnup LOCA experiments

The process for developing and qualifying nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and postirradiation examinations; and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes ~20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In situ measurements generally provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the US nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the US Nuclear Regulatory Commission (NRC) outlined in its research information letter several technical issues that the industry must address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heat-up phase. By leveraging these advanced in situ capabilities, this work used in situ data generated from a simulated LOCA to better understand high-temperature creep and its effect on Zircaloy balloon and rupture performance. Here, this work used the BISON fuel performance code to assess the high-temperature creep model predictions with in situ data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

BISON Validation to In-Situ Cladding Burst Test and High Burnup LOCA Experiments

The process to develop and qualify nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and post irradiation examinations (PIEs); and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes approximately 20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long and drawn-out process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in-situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In general, the in-situ measurements provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the United States nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the Nuclear Regulatory Commission, through their research information letter, outlined a number of technical issues for the industry to address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heatup phase. Leveraging these advanced in-situ capabilities, this work intends to use insitu data generated from a simulated LOCA in the Severe Accident Test Station at Oak Ridge National Laboratory to better understand high-temperature creep and its impact on Zircaloy balloon and rupture performance. This work used the BISON fuel performance code to compare BISON’s high-temperature creep model predictions to in-situ data and identify limitations and gaps within the model. Additionally, BISON will subsequently be used to simulate relevant high-burnup LOCA tests. The results will be analyzed and compared to the available post-test data in a manner consistent with the approach outlined in the Nuclear Regulatory Commission research information letter.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of grain size and porosity on cladding failure in high-burnup UO 2 : A sensitivity and uncertainty study

Isotopic taggants are being studied to aid in the provenance assessment of nuclear materials. However, these taggants must be selected such that they do not adversely affect fuel performance during normal operation or accident scenarios. Taggants are known to affect the fuel’s grain size and porosity. In the work described in this paper, the BISON fuel performance code was used to assess the potential effects of taggants (i.e., grain size and porosity) on fuel rod behavior and cladding failure during a high-burnup, large-break loss-of-coolant accident. Here, 281 individual fuel rods from the same reactor core were modeled for a sensitivity study, a parametric study, and uncertainty quantification. The cladding failure predictions often exhibited stochastic behavior. After additional study, it was found that the cladding failure model is highly sensitive to residual error inherent to numerical approximation solvers. Some strategies to mitigate this sensitivity are discussed. The study found no relationship between known taggant effects and cladding failure status. However, taggants were found to affect the time and location of failure in certain rods. In conclusion, future work to continue investigating and validating these findings is briefly discussed.

Doped UO 2↗