COMPARISON OF FATIGUE AND FRACTURE BEHAVIOR OF WELDED AND SEAMLESS PIPE STEEL IN GASEOUS HYDROGEN
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Agnani, Milan↗
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HYDROGEN PRESSURE CYLCING OF SUBSCALE PIPES TO SIMULATE FULL-SCALE TESTING OF TRANSMISSION PIPELINES (IPC2024-134173)
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San Marchi, Christopher William↗
Harvesting Subsea Water Motion to Provide Clean Electrical Power
This TEAMER project had three main objectives related to characterizing the performance of a vortex-induced vibration (VIV) marine energy harvesting device called the Whatever Input To Torsion (WITT). The first was to characterize the motion of the WITT device as it vibrates on top of a pipe in a natural tidal flow. The second was to measure how much electricity the WITT device generates when it is deployed on top of a pipe that is secured to the seabed with a lander. The third objective was to measure how much electricity the WITT device generates in the same configuration but with a longer pipe. This TEAMER project succeeded in satisfying the first objective by measuring the motion of the WITT device with an inertial measurement unit (IMU) mounted in the WITT housing. The motion was described by examining changes in the acceleration, pitch, roll, and yaw, and displacement. The frequency of the motion was characterized using spectral analysis. The TEAMER project succeeded in reaching the second objective by measuring the power produced by the WITT. It is reported here as the average and maximum power in hourly intervals. WITT Energy decided to not have us pursue the third objective of testing a longer pipe because the first test was unexpectedly successful at having the system vibrate at the target frequency. A longer pipe would have resulted in a decreased frequency and therefore not have been a useful test. Based on these findings, the decision was made to keep the same pipe length for both tests and improve on other aspects of the test structure that had failed during the first test. The key finding of this TEAMER project is that the WITT is capable of generating electricity in moderate flow speeds when it is in a housing that is attached to a 2.5 m pipe and secured to the seabed by a lander. This report contains the first power values for a VIV instrument that generates electricity from a pendulum swinging at the top of a long pipe. The experiment demonstrated that the vibration frequency of the device varied with the tidal current speed, which changed throughout a given tidal cycle. This meant that the desired peak frequency was not sustained for more than one hour. The success of this TEAMER test provides evidence that continued research in the area of electricity generation from VIV holds promise as a source of marine energy for offshore applications. Future tests should focus on refining the pipe length equations and improving the power take off system of the WITT power electronics. In addition, experiments aimed at expanded testing and evaluation (e.g. longer than one month) are needed to measure the power production at higher tidal current speeds and test the system's durability.
Branch, Ruth A. (ORCID:0000000202356719)↗
Determining Drying Conditions to Mitigate Hanford Transfer Line Corrosion
Radioactive waste is stored in underground, carbon-steel double-shell tanks at the Department of Energy Hanford site.1,2,3 The underground transfer lines are used to transfer the waste between the tanks and other assets at the Hanford tank farm facilities. The transfer lines constructed before 1990 are pipe-in-pipe design with most of the lines having carbon steel carrier pipe and carbon steel encasements. The lines constructed after 1990 have a stainless-steel carrier with a carbon steel encasement. The carrier lines, which are in contact with the waste, provide a critical confinement function, while the encasements provide secondary containment to reduce the risk of a release and contamination to the environment. In 2022, a visual inspection of the annulus between the encasement and the carrier for one of the pre-1990 transfer lines revealed standing liquid in contact with the carrier pipe. Additionally, a significant build-up of corrosion products on the exterior of the carrier pipe and the interior of the encasement pipe was observed. To assess the extent of condition, borescope inspections were performed in the encasement space of additional transfer lines in the facility to provide a baseline. The inspection revealed that of the lines inspected, 29 had either moisture present or evidence that moisture had been present in the past (e.g., waterline corrosion where the pipe had been exposed to standing water). Additional inspections indicated several instances of pitting corrosion on the interior of the encasement and exterior the carrier transfer lines. An image of the carrier pipe with several pits is presented in Figure 1(a) and the profiled image of the transfer line section along with the pit depths are presented Figure 1(b). As seen in the figure, the deepest pit was 73 mils on the pipe section. These lines were designated as “do-not-use” until further compensatory actions are implemented to return the lines to service.
Shukla, Pavan K. [Savannah River National Laborato↗
TuFF internal WRAP for Rapid Pipeline Repair (TuFF iWRAP)
The goal of “TuFF internal WRAP for Rapid Pipeline Repair” (TuFF iWRAP) program was to develop a novel material system and placement process to fabricate structural pipe within the existing deteriorated pipelines without disruption of gas delivery. The team (University of Delaware – Center for Composite Materials (UD-CCM) and Plitzie Inc.) addressed this challenge by developing a new material feedstock and pipe in pipe (PIP) repair strategy. This allows the potential for significant cost reduction and has minimum operational impact on gas customers. The new robotic based placement design allows discontinuous placement of pipe sections creating a stand-alone structural liner within the legacy pipeline without the need for pipe shutdown. Here, the material is supplied using a tethered material feeding system and is placed and UV cured with the internal Wound Rapid Automated Placement (iWRAP) system. This provides maximum placement efficiency capable of traversing 90 angle bends in 12-inch pipe and overall design customization to meet pipe repair requirements (e.g., variable wall thickness, bridging gaps, etc.). UV-curable fiber reinforced composite material has been optimized to meet structural performance and placement/cure times. Superior strength and fatigue life has been demonstrated by improving fiber-matrix adhesion using new fiber sizing for UV resins. Rapid cure approaches using new liner and resins have been evaluated with industry. The appropriate design of the section joints has been developed and tested. We estimate coating time to be ~100 hours per mile enabling typical pipe repair within 1 week.
03 NATURAL GAS↗
Leveraging MARVEL and SPHERE to Demonstrate NEAMS Thermal Hydraulics Codes
The Single Primary Heat Extraction and Removal Emulator (SPHERE) and Microreactor Applications Research Validation and Evaluation (MARVEL) were utilized to demonstrate thermal hydraulics codes of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. A recent long-duration test of a sodium heat pipe conducted at the SPHERE facility was utilized for the validation of the heat pipe code Sockeye. The first 10 hours of operation featured a startup of the heat pipe from room temperature; this portion was modeled with Sockeye and compared to the experimental data. Measurements included several thermocouple measurements across the length of the heat pipe surface. Two of Sockeye’s heat pipe models were applied to experiment, as well as several models of the cooling at the condenser end of the heat pipe. Sockeye results gave a decent match to the experimental data, with notable discrepancy in the condenser section, where experimental data indicated a significant inactive length of the heat pipe. The MARVEL microreactor was modeled through a coupled simulation of System Analysis Module (SAM) and SubChannel Module (SCM). This multiscale coupled simulation was performed as an exercise to demonstrate the compatibility and applicability of these codes for thermal-hydraulic analysis of microreactors. The coupling between the SCM and SAM codes for multiscale modeling of MARVEL, was achieved through a domain overlapping approach. In this example a transient SAM simulation provided boundary conditions to SCM, and SCM provided pressure drop to SAM.
97 - MATHEMATICS AND COMPUTING↗
Single Primary Heat Extraction and Removal Emulator (SPHERE) Long Duration Testing
For the development of heat-pipe cooled microreactors, it is crucial to thoroughly understand the characteristics and functioning of heat pipes across a wide spectrum of operating conditions. Passive heat removal and its long-term performance stability are critical factors in this context. Enhanced experimental data is vital for evaluating the operational lifespan of alkali metal heat pipes. Idaho National Laboratory (INL) has successfully conducted an extended duration test on a high-performance sodium-filled heat pipe, closely monitoring the axial temperature profile, power supplied by the heaters, and heat removed by a gas-gap calorimeter. The results from this testing provide valuable data that are instrumental in supporting heat pipe validation efforts. Specifically, this data aids in the development and validation of Sockeye, the Multiphysics Object-Oriented Simulation Environment (MOOSE) tool under the US-DOE NEAMS program designed for heat pipe modeling. By comparing experimental results with Sockeye’s predictions, the tool's accuracy and reliability can be assessed and improved, thereby enhancing its capability to simulate heat pipe operations under various conditions.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Single Primary Heat Extraction and Removal Emulator (SPHERE) Power Transient Testing
For heat pipe cooled microreactor development, it is essential to understand the characteristics of heat pipes and how they function under a wide range of operating conditions. Modeling the startup and shutdown of heat pipe cooled microreactors has been a major challenge for the modeling code. Validating these parameters within the code is key for the development and licensing of heat pipe cooled microreactor designs. Idaho National Laboratory (INL) has completed testing on power transients of a heat pipe with a range of operating conditions. The resulting temperature profiles from this testing can be utilized to aid in the validation of the startup and shutdown portions of the heat pipe modeling code.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Single Primary Heat Extraction and Removal Emulator (SPHERE) Long Duration Testing
For heat-pipe cooled microreactor development, it is essential to understand the characteristics of heat pipes and how they function under a wide range of operating conditions. An important process for heat-pipe cooled microreactor development is the passive heat removal and how the performance may change over long periods of time. Increased experimental data provides additional information to assess the operational lifetime of alkali metal heat pipes. Idaho National Laboratory has completed testing over a long duration of a high-performance sodium filled heat pipe while monitoring axial temperature profile, power supplied by the heaters, and heat removed by a gas-gap calorimeter. The results from this testing can aid in heat pipe validation efforts.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Data-Informed Synthetic Networks of Water Distribution Systems for Resilience Analysis in Puerto Rico
The increasing potential of infrastructure disruptions calls for high-quality infrastructure models to be used in resilience analysis and decision making. Unfortunately, many utilities and communities do not have access to accurate and detailed models due to a lack of data and resources. Furthermore, security restrictions on sharing infrastructure models present roadblocks to research, analysis, and decision making. Recent advances in the development of synthetic water distribution models provide a potential solution to this problem. There is an opportunity to improve these methods by leveraging incomplete pipe datasets to aid synthetic network generation. To address this gap, we developed a methodology for synthetic network generation that incorporates partial pipe data using a modification of the minimum cost flow algorithm for network generation and pipe sizing. This methodology demonstrates how partial pipe data can be leveraged to improve site-specific synthetic network generation. For the study area of Mayagüez, Puerto Rico, a synthetic model generated using 50% of real pipe data matches the pressure of the validation system with an average error of 23.5 m of head, which improves upon the average error of 31.6 m of head produced by a synthetic model generated using no data of the real pipes. Additionally, synthetic networks are shown to replicate the pressure response under a disruption scenario of the validation network, suggesting potential use in resilience analysis.
resilience analysis↗
DESIGN ISSUES AND QUALIFICATION OF HYDROFORMED DOUBLE WALLED EXPANSION JOINTS IN VACUUM SERVICE
The Vacuum Auxiliary System for ITER is devoted to pumping out, venting, and purging the vacuum volumes of the tokamak. Over 5000 clients including the cryostat and vacuum vessel, at 8500 m3 and 1400 m3 respectively, are serviced by approximately 150 pumping stations through 6 km of pipework. The piping and functions provided to the clients include component operation, routing of potentially tritiated gases, and timely leak localization. Expansion joints are used to reduce loadings on pumps and piping stresses but also to qualify in-line components having low allowable design loads. The ITER project utilizes vacuum valves that are not designed to withstand typical piping system loadings and require detailed design and evaluation. Loading scenarios prescribed by governing system specifications must be considered. To maintain allowable design loads for components and to keep piping stresses below code requirements, double walled hydroformed expansion joints have been employed in the design. These are not standard items and require custom fabrication. Installation space constraints and adjacent pipe support attachment location availability at ITER limit standard expansion joint installation guidelines provided by the Expasion Joint Manufacturers Association (EJMA). Careful consideration must be given to the analysis model to ensure proper function and life expectancy of the component. These considerations include accurate accounting of thrust forces, thermal movements, seismic accelerations, equipment and building differential displacements. In addition to displacements, the process internal, external, and interspace pressures affect the qualification and selection of the double walled expansion joints. The calculation results shall confirm that deflections, forces, and moments are reasonable for the size and type required for the system’s demands as evaluated against the manufacturer’s design.
Clark, Forrest [ORNL] (ORCID:0009000678106843)↗
Multiphysics Modeling of Microreactors with NEAMS codes, and Validation Based on KRUSTY Reactivity Insertion
The NEAMS Multiphysics Applications team continues to assess code usability and functionality for microreactor design and safety analyses, while demonstrating that NEAMS tools capture both steady-state and transient behavior across distinct microreactor concepts. In FY2025, the team advanced full-core, high-fidelity, multiphysics models that solve more complex problems and strengthen verification/validation for several microreactor systems: heat-pipe microreactor (HPMR), gas-cooled microreactor (GCMR), and the KRUSTY experiment. These models employ the MOOSE MultiApp/Transfers architecture with Griffin for neutronics, BISON for heat conduction/thermomechanics, Sockeye for heat pipes, SAM/THM for coolant channels and loops, and SWIFT for hydride behavior, with meshes generated via the MOOSE Reactor Module. The graphite models available in the Grizzly code were also investigated for future analyses. For the HPMR, a Na-HPMR variant was constructed to align with recently validated heat-pipe experiments and Sockeye’s LCVF capability, enabling mechanistic heat-pipe transients and startup modeling. The Na-HPMR will serve as the primary model for HPMR investigations in upcoming tasks. The load-following and single heat-pipe failure scenarios (Griffin/BISON/Sockeye), which were previously modeled for the K-HPMR, were replicated for the Na-HPMR, showing strong negative temperature feedback and highly localized thermal effects, respectively, while the startup case captured vapor-front progression and heat-removal activation. Solid mechanics was added to the previously built K-HPMR full-core model in BISON, showing minimal impact on steady-state reactivity yet enabling stress-field predictions that prepare the path for full-core TRISO performance analyses. For the GCMR, automated steady-state and four transient scenarios were executed using Griffin/BISON/SAM/SWIFT. Results confirm robust inherent safety: power collapses promptly in loss-of-cooling events, the inlet-temperature drop settles to a new equilibrium, and a single-channel blockage yields only a ~30 K local fuel-temperature rise with <0.4% power decrease. SWIFT-predicted hydrogen redistribution affects reactivity during both steady-state and transient conditions, underscoring its importance. A Brayton-cycle balance of plant (BOP) model in SAM/THM demonstrated stable startup behavior, and xenon-driven reactivity during load following was analyzed. To improve TRISO-compact temperature fidelity, a fast multiscale Heat Source Decomposition (HSD) treatment was implemented. Against heterogeneous benchmarks, HSD reduces underprediction of kernel temperatures and lowers predicted peak powers in reactivity-insertion transients compared to previous homogenized models. KRUSTY warm-critical validation progressed from FY2024 baselines: the 15Ȼ insertion shows excellent agreement in peak power (~2% high) and temperature trends, and the 30Ȼ case was automated via a feedback controller that maintained power near 3 kW for ~150 s with close agreement to data. The successful modeling of the warm critical tests has laid a strong foundation for simulating more complex nuclear system tests in the years ahead. Throughout FY2025, developer feedback was provided (e.g., MOOSE batch mesh generation, distributed pre-split meshes, Griffin sweeper on displaced meshes), several new models were contributed to the Virtual Test Bed, and an OECD-NEA WPRS multiphysics benchmark based on the HPMR was initiated to enable broader cross-comparison and best-practice development with the nuclear community at large.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Experimental Validation of Exact Burst Pressure Solutions for Thick-Walled Cylindrical Pressure Vessels
Burst pressure is one of the critical strength parameters used in the design and operation of pressure vessels because it represents the maximum pressure that a vessel can withstand before failing. Historically, the Barlow formula was used as a design base for estimating burst pressure. However, it does not consider the plastic flow response for ductile steels and is applicable only to thin-walled cylinders (i.e., the diameter to thickness ratio D/t ≥ 20). A new multiaxial plastic yield theory was developed to consider the plastic flow response, and the associated theoretical (i.e., Zhu–Leis) solution of burst pressure was obtained and has gained extensive applications in the pipeline industry because it was validated by different full-scale burst test datasets for large-diameter, thin-walled pipelines in a variety of steel grades from Grade B to X120. The Zhu–Leis flow theory of plasticity was recently extended to thick-walled pressure vessels, and the associated exact flow solution of burst pressure was obtained and is applicable to both thin and thick-walled cylindrical shells. Many full-scale burst tests are available for thin-walled line pipes in the pipeline industry, but limited pressure burst tests exist for thick-walled vessels. To validate the newly developed exact solutions of burst pressure for thick-walled cylinders, this paper conducts a series of burst pressure tests on small-diameter, thick-walled pipes. In particular, six burst tests are carried out for three thick-walled pipes in Grade B carbon steel. These pipes have a nominal diameter of 2.375 inches (60.33 mm) and three nominal wall thicknesses of 0.154, 0.218, and 0.344 inches (3.91, 5.54, and 8.74 mm), leading to D/t = 15.4, 10.9, and 6.9, respectively. With the burst test data, comparisons show that the Zhu–Leis flow solution of burst pressure matches well the burst test data for thick-walled pipes. Thus, these burst tests validate the accuracy of the Zhu–Leis flow solution of burst pressure for thick-walled cylindrical vessels.
42 ENGINEERING↗
Moderated nuclear reactor
A heat pipe cooled nuclear reactor core (“reactor core”) includes an array of fuel, an array of primary heat rejection heat pipes, and an array of moderator cooling heat pipes or thermosiphons. Each moderator cooling heat pipe in the array of moderator cooling heat pipes is surrounded by metal hydride moderator materials, slowing neutrons from a fission energy range to a lower energy range in the reactor core.
Reid, Robert Stowers↗
HYDRODYNAMIC TESTING OF ZIRCALOY TO STEEL HUB COUPLINGS FOR USE IN ADVANCED TEST REACTOR I-LOOP TUBE
The ATR I-positions' isolation from the flux trap region provides an opportunity for a test facility that could create conditions representative of commercial light water reactors. Such a facility requires piping couplings and joints that withstand extreme thermal and pressure cycling. Also, transitioning from steel to zircaloy piping inside the reactor vessel allows the neutronic properties of zircaloy to be leveraged while using cost-effective stainless steel outside of the core. Grayloc-style hub couplings will act as transition points between stainless steel piping and zircaloy piping inside the reactor vessel. These couplings have not been sufficiently tested for leakage where there is a transition between materials. This transition is a concern because the thermal expansion coefficients of the materials differ. Seal integrity between the two components may be greatly affected by thermal expansion at the connection. Thus, the effect of thermal cycling must be assessed. Thermal and pressure cycling and hydrodynamic testing will be conducted to verify seal integrity of the steel to zircaloy hub coupling when subjected to ATR conditions.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Quantifying uncertainty in uranium concentration measurements via K-edge densitometry
This study quantifies the uncertainty in uranium concentration predictions of fluoride and chloride-based salts within a steel pipe using K-edge densitometry. Modeling and simulation was conducted with the Monte Carlo N-Particle Transport (MCNP) code. The quality of of this technique’s prediction in a pipe requires proper characterization of the pipe’s thickness, which is dependent on the source size and axial offset from the pipe centerline. The thickness was determined as either the center-line thickness seen by the X-ray source or an average value determined through random sampling. Generally, the predicted concentrations were slightly better at lower offset with the random sampling thickness and using the center-line thickness for the highest offsets. For a line-beam source and varying axial offsets, the relative error of concentration was within 1% of the true value but uncertainty increased by 2 orders of magnitude. Similarly, for no axial offset, the relative error was significantly less than 1% while no trend for uncertainty was found. However, at the largest possible offset for a given source size, the concentrations become erroneous and greater than the allowable 1% relative error. Furthermore, high offsets tended to increase the variance of the transmission spectra by 3 orders of magnitude.
Characterization and Analytical Technique↗
Special Nuclear Material Hold-up Measurement
Facilities that process special nuclear material (SNM) generally have a variety of equipment, shielding, and nearby radiation sources, which can pose challenges when performing nondestructive analysis (NDA) within the facility, as shown in Figure 1. In this figure, the left image features two individuals performing a holdup measurement of a large duct at a posting, utilizing a detector probe attached to a yardstick to reduce some aspect of the measurement uncertainty. Note that the distance to the holdup within the pipe from the posting cannot be accounted for. The right image similarly features several individuals guiding a holdup measurement of a ducting pipe several meters overhead, by attaching a detector probe to a long handle and lining that up with a posting on the outside of the pipe. SNM processing begets SNM holdup, and this difficulty with accurate NDA can result in unaccounted accumulation of SNM holdup within a process or area. Normal SNM processing operations can accumulate holdup within filters, pumps, pipes, ducts, other equipment, and facility support systems, as demonstrated by Figure 2. It is important for deactivation and decommissioning (D&D) gloveboxes and their ancillary equipment, material accountability, criticality safety, facility operations, waste management, radiation safety, and security to accurately characterize the location, composition, and quantity of this holdup to minimize these associated risks and potential diversion pathways. Current NDA techniques for assay of holdup are expensive, time consuming, and have high uncertainty.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Advancing 1D thermo-hydraulic tools for large cryogenic facilities
The Cryogenic Division at Fermilab develops large-scale cryogenic systems for particle accelerators and superconducting test facilities. To support design and diagnostics, a Python-based code was created to calculate pressure drops in components such as valves and pipes. This paper presents recent enhancements to the code, including new heat transfer functions that improve the accuracy of thermal and hydraulic predictions. The first application models pressure and temperature changes in PIP-II relief pipes, aiding pipe sizing and protecting relief valves. The second example analyzes heat load evolution in a pipe carrying sub-atmospheric helium, helping interpret temperature sensor data and understand gas return behavior to cold compressors. These improvements significantly expand the tool’s capabilities, offering a practical resource for designing and troubleshooting cryogenic systems under dynamic thermal and flow conditions.
Beckwith, Rosalyn [Fermilab]↗