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Highlights, milestones completed, major accomplishments/activities, upcoming meetings/events, awarded research projects, cooperative research projects, and infrastructure management initiative are covered in this November 2014 report of the Advanced Test Reactor national Scientific User Facility.
Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of Cambridge have been investigated. The rationale for the superior performance of the type N using a customized sheath developed at the University of Cambridge is explained in comparison with the behavior of conventional type N Inconel 600 sheathed thermocouples. (authors)
This is a powerpoint presentation for the annual Advanced Sensors and Instrumentation (ASI) program webinar. The following is a summary of the presentation. Advanced instrumentation enables testing of nuclear fuels and materials in support of the US advanced nuclear technology industry. A number of promising sensor technologies are in the pipeline. The early part of sensor development can be done outside of the reactor environment, but full technical readiness requires experience gained from in-core performance testing. Customers usually have only one shot to conduct their irradiation experiments. Because of the high costs test sponsors are frequently reluctant to incorporate unproven technologies in reactor experiments. Therefore, it is vital to demonstrate newly-developed sensors in operational conditions, prior to incorporating them into long-term high-value experiments. Successful completion of these activities will create new monitoring capabilities in high-power test reactors (such as ATR) as well as specialty reactors such as TREAT.
This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR-2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. 2. Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tristructural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S.-produced fuel.
This fact sheet will provide information for the database of experimental facilities supported by the IAEA as outlined below: NEXSHARE: As part of the Nuclear Harmonization and Standardization Initiative (NHSI), the IAEA proposes to establish a Network for global cooperation and resource sharing for experiments and code validation between experimental facilities, SMRs design organizations, International Organizations and Technical Support Organizations (TSOs). The proposed Network, NEXSHARE, will be done in collaboration with the OECD/NEA. EXPERIMENTAL FACILITY DATABASE: As part of this activity, the IAEA is compiling a database of experimental facilities applicable to SMRs (including water cooled, high temperature gas cooled, molten salt and fast neutron spectrum reactors). This database will also form part of NEXSHARE and will be documented in an IAEA publication.
The gamma-ray emitting radionuclide concentration and decontamination factor results from gamma-ray spectrometry measurements of ATR loop liquid samples by the Radiation Measurements Laboratory (RML) are contained in this report for ATR Cycle 175D from Jan. 23, 2026, to Apr. 26, 2026.
The Advanced Test Reactor’s (ATR’s) distinctive ability to provide a wide range of irradiation conditions is attractive for programs pursuing fuel qualification experiments. These potentially high-fuel-load experiments are a relatively new development and produce unexplored effects on nearby experiments. Here, this paper explores how photon heating of such an experiment may affect other nearby experiment programs, ultimately serving to better inform decisions regarding experiment design and risks to programmatic goals. The MC21 (Monte Carlo for the 21st Century) code is used to model and study how gamma heat generation rates and axial effects impact different ATR positions. The results reveal that the proximity of a given experiment’s position to the high-fuel-load one can significantly alter that experiment’s expected axial profile.
Highlights: Mike Worley and Shane Johnson visited INL Jan. 22 for an NSUF strategy discussion; Rory Kennedy attended a NSLS-2 Beamline Advisory Team meeting at Brookhaven; Provided a final cost estimate to the NSUF Program Office in support of the NEET/NSUF proposal, “Metal-ceramic and metal-metal composites for extreme radiation and temperature environment: An in situ interface stability and mechanical behavior study by high energy x-ray diffraction with a synchrotron probe.”; Assisted in the development of conceptual designs and performed a preliminary thermal hydraulic analysis for two NEET/NSUF proposals. The challenge for both experiments is to provide high (>1000 C and up to 1600 C)) specimen temperatures in a small space (0.5" diameter ATR Outboard A-position) without overheating the coolant. Several designs were analyzed and found to be feasible, although detailed design and analysis will be required after the projects are awarded; and A single USU TEM specimen is packaged and awaiting shipment from MFC to CAES. Once at CAES, SEM, TEM and LEAP analysis will be performed. Professor Ban has requested additional sub-samples to be made to take back to his laboratory at USU for thermal diffusivity studies.
The success of modular high temperature gas-cooled reactors is highly dependent on the performance of the tristructural-isotopic (TRISO) coated fuel particle and the quality to which it can be manufactured. During irradiation, TRISO-coated fuel particles act as a pressure vessel to contain fission gas and mitigate the diffusion of fission products to the coolant boundary. The fuel specifications place limits on key attributes to minimize fuel particle failure under irradiation and postulated accident conditions. PARFUME (an integrated mechanistic coated particle fuel performance code developed at the Idaho National Laboratory) was used to calculate fuel particle failure probabilities. By systematically varying key TRISO-coated particle attributes, failure probability functions were developed to understand how each attribute contributes to fuel particle failure. Critical manufacturing limits were calculated for the key attributes of a low enriched TRISO-coated nuclear fuel particle with a kernel diameter of 425 μm. As a result, these critical manufacturing limits identify ranges beyond where an increase in fuel particle failure probability is expected to occur.
We report as a system level code, RELAP5 is widely used in the nuclear field for the reactor hydraulic analysis. In the component/experimental level, it is frequently employed as well. This paper demonstrates that the RELAP5 flow simulation of the AFIP-7 irradiation experiment which has a complicated 3D geometry/flow path deviates by ~28% from the flow evaluation through a computational fluid dynamics (CFD) simulation, which has been validated against a flow test performed at Oregon State University. Thermal safety compliance analysis of an experiment planned to be irradiated in the Advanced Test Reactor is usually performed using the finite element analysis code, Abaqus. This research reveals the disadvantages of the Abaqus simulation in the flow instability and departure from nucleate boiling evaluations via a conjugate heat transfer analysis of the AFIP-7 irradiation experiment. As a result, a detailed CFD simulation is suggested for irradiation experiments with complicated flow paths, rather than a simplified RELAP5 simulation or hand calculations. The validated simulation approach will be integrated with the Boehmite correlations to investigate the oxide growth prediction on the fuel cladding in Part II of this research.
This study explores data-driven anomaly detection methods to analyze sensor fail- ures in the Advanced Gas Reactor (AGR) nuclear fuel irradiation experiments. Specifically, we examine failures of thermocouples (TCs), which are critical for mon- itoring and controlling in-reactor temperatures during operation. Failures were pri- marily observed during abrupt power transitions and manifested as sensor drop-outs, drifts, or unexplained behavior. We applied three time-series analysis techniques— rolling mean smoothing, matrix profile, and vector auto-regression (VAR)—to de- tect anomalies in TC data prior to failure events. The rolling mean method effec- tively highlighted deviations aligned with reported failures, while the matrix profile provided partial early warning but sometimes flagged normal fluctuations during power-down periods. VAR shows potential in capturing multivariate dependencies but requires further calibration. A rare case of TC drift was also documented, which did not result in failure, underscoring the challenge of building predictive models with sparse positive examples. Our findings demonstrate that traditional statistical tools can aid anomaly detection but have limited predictive power without richer training data. We propose future directions including synthetic data generation, real- time surrogate modeling, and multi-modal feature integration. This work provides a foundation for applying robust anomaly detection frameworks to mission-critical sensor systems in experimental settings.
A temperature sensitivity evaluation has been performed for the AGR-2 fuel experiment on an individual capsule. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameters are the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. Thermal conductivity of the fuel compacts and thermal conductivity of the graphite holder were in the middle of the list for sensitivity. The smallest effects were for the emissivities of the stainless steel and graphite, along with gamma heat rate in the non-fueled components. Sensitivity calculations were also performed varying with fast neutron fluence, showing a general temperature rise with an increase in fast neutron fluence. This is a result of the control gas gap becoming larger due to the graphite holder shrinkage with neutron damage. A smaller sensitivity is due to the thermal conductivity of the fuel compacts and graphite holder decreasing with fast neutron fluence.
In support of the water conservation goals of the U.S. Department of Energy’s Strategic Sustainability Performance Plan, a number of water conservation projects have been proposed that would reduce the volume of wastewater currently discharged to the Advanced Test Reactor Complex’s Cold Waste Pond at the Idaho National Laboratory. This evaluation provides an approximation of the potential impact to groundwater quality from reducing raw or “clean” water discharges to the Cold Waste Pond and concentrating chemicals (e.g., sulfate) in the effluent.
PNNL has completed sectioning of the U.C. Berkeley hydride fuel rodlet 1 (highest burn-up) and is currently polishing samples in preparation for optical metallography. A disk was successfully sectioned from rodlet 1 at the location of the internal thermocouple tip as desired. The transition from annular pellet to solid pellet is verified by the eutectic-filled inner cavity located on the back face of this disk (top left) and the solid front face (bottom left). Preliminary low-resolution images indicate interesting sample characteristics in the eutectic surrounding the rodlet at the location of the outer thermocouple tip (right). This sample has been potted and is currently being polished for high-resolution optical microscopy and subsequent SEM analysis.
The Nuclear Science User Facilities (NSUF) and its partner facilities represent a prototype laboratory that utilizes a distributed partnership, with each facility bringing exceptional capabilities to the relationship including reactors, beamlines, state-of-the-art instruments, hot cells, and equally important, expertise. These capabilities and people together create a nationwide infrastructure that allows the best ideas to be tested using the most advanced capabilities. Through the NSUF, researchers and their collaborators are building on current knowledge to better understand the complex behavior of materials and fuels in a nuclear reactor. The NSUF’s partnership program in 2016 included eight universities, one research and education consortium, three national laboratories, and one industry partner. The avenues opened through these partnerships facilitate cooperative research across the country, matching people with capabilities and students with mentors. The NSUF in 2016 included Idaho National Laboratory (INL) and the following institutions: • Argonne National Laboratory • Center for Advanced Energy Studies (a research and education consortium between Boise State University, INL, Idaho State University, University of Idaho, and University of Wyoming) • Illinois Institute of Technology • Massachusetts Institute of Technology • North Carolina State University • Oak Ridge National Laboratory • Pacific Northwest National Laboratory • Purdue University • University of California, Berkeley • University of Michigan • University of Nevada, Las Vegas • University of Wisconsin • Westinghouse Materials Center of Excellence. This report contains details on new capabilities brought into the NSUF in 2016, the evidence of the increasing research facilitated by the NSUF, and new tools and activities to propel nuclear research and development forward to meet energy reliability and security needs. Learn about the NSUF Nuclear Fuels and Materials Library and the Nuclear Energy Infrastructure Database, and how the robust NSUF website continues to evolve to support new tools. Read about how the NSUF supports Department of Energy (DOE) missions through efforts like the 2016 NSUF Ion Beam Options Workshop, which led to a prioritiza-tion of ion beam facilities according to their ability to support Department of Energy Office of Nuclear Energy (DOE-NE) missions.
This engineering calculations and analysis report documents the specimen position/elevation adjustments from the third capsule of the Advanced Graphite Creep experiment (AGC-3). These adjustments are necessary due to irradiation induced shrinkage and load induced creep. These effects are calculated and integrated into a table that lists the specimen IDs, their stack number, their nominal elevation, and the average specimen elevation for each reactor cycle.
In the pursuit of the ability to perform multiphysics simulations of the Advanced Test Reactor, it is crucial to have a fast and highly accurate deterministic model. To achieve this, a contemporary two-step method is employed. The first step involves generating homogenized cross sections using OpenMC, a cutting-edge Monte Carlo neutron transport code. OpenMC offers excellent modular capabilities, allowing for easy component integration and flexibility in incorporating new designs into the model. The second step involves deterministic transport calculations, which are performed using Griffin, a reactor multiphysics application based on the Multiphysics Object-Oriented Simulation Environment. To ensure the accurate spatial resolution and assignment of material cross sections, a Cubit-generated mesh for the Advanced Test Reactor is utilized as an intermediate step between the OpenMC and Griffin models; Griffin utilizes the mesh for its finite element solution, while OpenMC material IDs are written to the mesh file to be used in Griffin material assignments. Additionally, a Python-based script converts the cross sections generated by OpenMC into the ISOXML format required by Griffin. Preliminary comparisons indicate good agreement between the neutron multiplication factors obtained from the standalone OpenMC model and the Griffin model, with differences of less than 50 pcm in the two-dimensional geometry configuration. However, in three-dimensional calculations, an unacceptably large error is found in the Griffin solution. Future work is planned to resolve this discrepancy.
The University of Tennessee, Knoxville and Oak Ridge National Laboratory have completed proof-of principle testing on an advanced sulfur-based zirconium and aluminum chlorination process for the recovery of Zr or Al from spent fuel claddings and purification of the products. This process could be used to recover hafnium-free nuclear-grade zirconium for reuse or disposal as low-level waste. Testing showed that even non-optimized reaction protocols could produce zirconium product streams from simulated irradiated fuel with purities >90%. The chlorination process was tested on representative fission product species that would be found in Advanced Test Reactor fuel (ATR). First, fission product generation was determined through modeling and simulation of the ATR. The generation of fission products was used to selected contaminants and potentially volatile fission product simulants for laboratory testing. The goal is to identify which species may react with the sulfur-chlorine reagents and be transported into the advanced chlorination off-gas.