The Advanced Test Reactor (ATR) relies on In-Pile Tubes (IPTs) to support a wide range of irradiation experiments. Existing IPT designs utilize stainless steel (SS348) structural components that introduce parasitic neutron absorption and reduce thermal neutron availability for irradiation experiments. This study evaluates the neutronic benefits of replacing stainless steel IPT components with zirconium-based alloys to enhance irradiation performance while maintaining reactor operating characteristics. A series of parametric analyses was performed using the validated full-core three-dimensional MC21 model developed for the ATR Cycle 175A Core Physics Safety Analysis. Two candidate configurations were examined: (1) replacement of the SS348 flow tube with Zircaloy-4, and (2) replacement of the flow tube with Zircaloy-4 and the pressure and envelope tubes with Zr-2.5Nb. Thermal neutron, fast neutron, and photon fluxes were evaluated within a representative IPT location, and impacts on reactor operating parameters were assessed. Results show that replacing only the flow tube with Zircaloy-4 increases incident thermal neutron flux by approximately 10% relative to the current stainless-steel configuration. Replacing the flow tube, pressure tube, and envelope tube with zirconium-based alloys increases thermal neutron flux by approximately 86% within the central irradiation region. In both cases, fast neutron and photon environments remain comparable to baseline conditions, and predicted operating characteristics, including outer shim control cylinder and neck shim behavior, exhibit negligible changes. The analysis demonstrates that zirconium-based SIPT components provide a practical, low-risk approach for substantially increasing thermal neutron availability in ATR in-pile tubes. The fully zirconium-based configuration offers the greatest benefit and could significantly improve experiment performance and irradiation flexibility without requiring corresponding increases in reactor lobe power.

Nuclear power is experiencing a generational resurgence, creating new opportunities in grid electricity, remote communities, behind-the-meter deployment, microgrids, industrial-process integration, and other applications. The University of Illinois Urbana-Champaign, through its Department of Nuclear, Plasma and Radiological Engineering, plans to deploy Nano Nuclear Energy's 45 MWth KRONOS MMR technology as a microreactor demonstration project. The project will create a research facility focused on advancing knowledge, building operational experience, and supporting technology development. On March 31, 2026, the university submitted an application to the U.S. Nuclear Regulatory Commission for a construction permit for the campus research reactor. On May 18, 2026, the NRC informed the university that staff had sufficient information to begin detailed safety and environmental reviews and establish a predictable review schedule. The presentation provides a design overview and an update on application review and the regulatory-audit plan and schedule.

The Advanced Test Reactor Critical (ATR-C), at the Idaho National Laboratory (INL), is a full -scale replica of the core of the Advanced Test Reactor (ATR) but located in an open pool in lieu of a pressure vessel. While ATR can operate at a nominal thermal power capacity of 250 MW, ATR-C is typically operated at less than 600 watts. ATR-C is most often utilized to evaluate experiment impacts on power distribution prior to insertion into ATR. The Radiation Measurement Laboratory (RML) performs measurements on a minimum of 340 uranium-aluminum fission wires that are distributed within fuel elements in the core during the irradiation to assist in validating modeling and simulation results. As these fission wires decay rapidly, all must be measured within a 12-hour window of the reactor scram. ATR-C is also used for general nuclear research purposes to demonstrate feasibility before investing in an irradiation campaign in ATR. Utilizing guidance provided in several ASTM International standards on neutron dosimetry, material radioactivation has taken place in a series of test positions within the core as well as in instrument thimbles, dry, air-filled positions located just outside the main reactor tank. Gamma spectroscopy measurements of the activated material have yielded results that, when used in conjunction with the methods within the ASTM standards, characterize fast and thermal neutron flux rates in these assorted test positions.

Presented by Christine Norman (University of Missouri) on behalf of Ethan Schroeder.

Neutron-beam science is a key functional focus of the NextGen MURR design, intended to strengthen domestic capability for materials science, energy efficiency, nondestructive testing, biology, and medicine. The pre-conceptual site layout includes a neutron-guide hall capable of hosting 10 to 15 instruments, a cold source, and two cold and two thermal neutron beam ports. In July 2026, MURR and NextGen MURR representatives attended the American Conference on Neutron Scattering. MURR organized a University-Based Neutron Sources Workshop to foster collaboration and address national neutron- access needs. Participants included universities, national laboratories, international organizations, the U.S. Department of Energy, and the Neutron Scattering Society of America. Priority instruments identified included small-angle neutron scattering, neutron imaging, neutron powder diffraction, and triple-axis spectroscopy. A beam-science expert panel will continue advising the design team.

Accurate characterization of displacement damage produced by fast neutrons is essential for qualifying semiconductor devices for aerospace, defense, fusion, and nuclear-energy environments. ASTM E1855 provides a standardized method for expressing neutron displacement damage in silicon as an equivalent fluence of one-MeV neutrons. This work applies ASTM E1855 to benchmark the University of Massachusetts Lowell Research Reactor Fast Neutron Irradiator. Commercial 2N2222A bipolar-junction transistors were electrically characterized before irradiation, irradiated at Sandia National Laboratories' Annular Core Research Reactor to approximately 1 x 10^13 n/cm2 one-MeV silicon-equivalent fluence, and characterized after irradiation to establish calibration response. Calibrated devices were then irradiated in the UMLRR Fast Neutron Irradiator to determine its equivalent one-MeV silicon damage rate and compare performance with the ACRR reference field. The benchmark establishes traceability to a nationally recognized reference neutron field and supports consistent semiconductor displacement-damage testing across university reactors and accelerator-based neutron sources.

In March 2026, Idaho State University, in partnership with NuCube Energy Inc., was accepted into the U.S. Department of Energy Reactor Pilot Program and subsequently entered the NRIC Launch Pad USA program. The application represented several firsts: the first university-led application and the first public-private partnership involving a state entity, along with challenges associated with those first occurrences. The presentation discusses development of the relationship, the intended roles and outcomes of each entity, and shared goals for eventual licensure of both a commercial product for national energy production and an institutional resource for research, testing, and training.

The Advanced Test Reactor Critical (ATRC) Facility is used to support nuclear instrumentation calibration, support for experiment insertion into the Advanced Test Reactor (ATR), reactivity measurements, void reactivity measurements, and axial profile measurements. New experiments scheduled to be inserted into ATR operating cycles may need to be tested in the ATRC first, either with a mockup or the prime experiment, to accurately capture flux profiles and reactivity worths. This information is necessary for authoring ATR Core Physics Analyses (CPA) and Core Safety Assurance Packages (CSAP). This process requires developing a separate test plan and ATRC operations which can extend the critical path of ATR operations or lead to the experiment being removed from the cycle. Validating the full fidelity Monte Carlo neutronics model against measured ATRC experimental data can reduce the number of experiment measurements required in the ATRC prior to irradiation in the ATR. This would take the requirement of ATRC testing off ATR's critical operating path and allow for increased flexibility of the ATRC to be used as a valuable tool, on its own, for low power tests. This validation looks at the neutronics model of the ATRC, previously tested experiments and their models, in all the testing positions that influence the flux and power distribution in ATR physics safety analyses.

Abilene Christian University's Nuclear Energy eXperimental Testing Lab is designing, licensing, and constructing the Molten Salt Research Reactor, a one-megawatt-thermal liquid-fueled molten-salt reactor. The NEXT Lab mission is to provide solutions for energy, water, and medical-isotope needs by advancing molten-salt-reactor technology while educating future nuclear-science and engineering leaders. The reactor is sponsored by Natura Resources and developed with The University of Texas, Texas A&M University, and Georgia Institute of Technology. ACU has established research and development laboratories for molten-salt chemistry, materials testing, and chemical analysis. The U.S. Nuclear Regulatory Commission issued ACU a construction permit authorizing reactor construction within the Dillard Science and Engineering Research Center in Abilene, Texas. The presentation provides an overview of the NEXT Lab mission, the reactor project, the collaborative team, and the status of development and licensing.

Flux determination is important both for experimental measurements and for assessing reactor condition. At the TRIGA Mark II reactor at NETL, the thermal pneumatic neutron irradiation station is expected to have a predominantly thermal spectrum. Three dosimetry wires were irradiated with a Xe-134 sample as flux monitors for measurement of the isomeric yield ratio of Xe-134(n,gamma)Xe-135. Flux values were determined through spectral adjustment using STAYSL PNNL to adjust a previously generated SAND-II spectrum by a least-squares approach. The method reconciles measured wire activation with a spectrum based on prior wire activations. The measured thermal flux was 2.69 x 10^12 n/cm2-s, the epithermal flux was 6.97 x 10^11 n/cm2-s, and the fast flux was 7.76 x 10^11 n/cm2-s, with associated uncertainties.

Presented by Michael Hoehn on behalf of Ethan Schroeder.

The University of Missouri is undertaking a transformational initiative to design, license, and construct a next-generation, higher-power research reactor known as NextGen MURR. The pre-conceptual design is envisioned to operate at 20 MW thermal or greater, double the power of MURR, using an open tank-in-pool configuration and high-density U3Si2 low- enriched uranium fuel enriched to 19.75%. The design demonstrates world-class perturbed thermal-neutron flux greater than 5.0 x 10^14 n/cm2-s, supporting production of critical medical and industrial radioisotopes. The site layout envisions an integrated nuclear ecosystem supporting isotope processing, neutron-beam science, radiopharmaceutical research, materials research, and workforce development at Discovery Ridge Research Park in Columbia, Missouri. The project will leverage university, industry, design-team, and owner's-representative expertise to establish a world- class center for nuclear excellence capable of adapting to state and national nuclear-science and technology needs.

Experiments developing a direct-energy-conversion system are being planned using the TRIGA Mark II research reactor at The University of Texas at Austin Nuclear Engineering Teaching Laboratory. A test device will be placed in Beam Port 1 adjacent to the core to provide neutron and gamma-ray doses that enable operation. Accurate knowledge of the radiation source affecting the device is crucial to assessing performance. This work characterized neutron flux as a function of energy and position in the Beam Port 1 core-adjacent position through experimental activation-foil measurements and spectrum unfolding with the STAYSL code package, together with computational characterization using Monte Carlo simulation of the physical experiment. The neutron-flux results were compared with sample activation in the beam port to inform experimental performance and radiological safety..

Deployable Energy's Unity Zero Power Critical Experiment was the third advanced-reactor demonstration to achieve criticality during the year in the Neutron Radiography Reactor facility. Integrating an external critical-experiment platform into an operating research-reactor facility provided operations and maintenance insights. The project required outage planning, a safety-analysis-report update, and electrical, mechanical, and instrumentation- and-control modifications to accommodate the Unity assembly. Procedure development integrated Unity operational requirements with established NRAD processes, while operator training combined classroom instruction, system walkthroughs, and simulated approach-to-critical activities. Initial criticality was achieved June 30 at 11:55 p.m. The reactor demonstrated stable and predictable behavior, confirming modeling of a well-characterized fuel type. The project provides lessons in equipment integration, procedure writing, training, and maintenance planning for facilities considering similar demonstrations.

The Advanced Test Reactor (ATR), at Idaho National Laboratory, is a highly versatile research reactor utilized for a wide variety of nuclear research applications. Its unique clover-leaf core design and adjustable outer shim control cylinders allow for the neutron flux to be controlled within each of the nine flux traps thereby facilitating simultaneous experiment irradiations with different exposure requirements. Due to the heterogenous neutron flux profile within the ATR core, dosimetry irradiation and subsequent measurement is essential to validate experiment exposure during an irradiation cycle. The Radiation Measurement Laboratory (RML) performs these measurements, and many more, in support of the ATR mission. Until recently, several of these measurement processes remained unchanged since their initial development. Failures of custom made, legacy hardware and software complications at RML illustrated the pressing need for modernization of equipment and methodology. Efforts are underway to align measurement practices with current industry standards, replace legacy Fortran analysis software, and upgrade or replace custom-built equipment with off-the-shelf commercial hardware all to improve program resiliency and measurement quality.

Eden Radioisotopes is a startup company based in Albuquerque, New Mexico. Eden's plan is to build an integrated reactor and hot cell facility in southeastern New Mexico dedicated to the commercial production of medical radioisotopes for the US. The Eden reactor is a 2-MW pool-type reactor called the All-Target Reactor, licensed from Sandia National Laboratories (SNL). The All-Target Reactor uses low-enriched uranium (LEU) fuel targets in the core to produce fission molybdenum-99 (Mo-99) and other medical radioisotopes. All of the targets are processed to make Mo-99. There are no driver fuel elements in the core, and hence there is no spent fuel produced by operating the reactor. The reactor also has several irradiation locations used for the production of activation medical radioisotopes such as Lutetium-177 (Lu-177) and others. In June 2026, the Nuclear Regulatory Commission (NRC) docketed the Eden Preliminary Safety Analysis Report (PSAR) and began the formal review process of the Construction Permit Application (CPA). When the construction permit is issued, Eden will begin construction of the integrated reactor and hot cell facility. Eden plans to be operational by 2030. When in full production, Eden will be able to produce 100% of the US demand for Mo-99 and the world demand, if needed. Eden will also be capable of producing a significant quantity of Lu-177 and other activation isotopes to meet the world demand. This paper will focus on the design features of the integrated reactor and hot cell facility, called the Eden Medical Isotope Facility (EMIF) and the All-Target Reactor.

A simulated gas source was created for efficiency calibration at the TRIGA reactor at The University of Texas at Austin Nuclear Engineering Teaching Laboratory. The facility irradiates gases whose distributed geometry differs from a conventional button-source geometry. Natural europium was placed in a canister with beads to create a gas-like distribution and irradiated to produce europium-152. Standard energy lines from 122 keV to 1408 keV were used to calculate efficiency values. A relative efficiency value was calculated for the xenon-135m 526.6-keV line and xenon-135 249.8-keV line; the relative efficiency was 0.5933 +/- 0.0003.

The University of Missouri is undertaking a transformational initiative to design, license, and construct a next-generation, higher-power research reactor known as NextGen MURR. The pre-conceptual reactor is envisioned to operate at 20 MW thermal or greater using an open tank-in-pool configuration and high-density U3Si2 low-enriched uranium fuel enriched to 19.75%. The design demonstrates perturbed thermal-neutron flux greater than 5.0 x 10^14 n/cm2-s, supporting production of medical and industrial radioisotopes. Integration with a co-located Radioisotope Processing Facility will support efficient handling and processing of irradiated targets. The modular and scalable RPF and support utilities allow expansion as demand increases. The site layout also includes infrastructure for neutron research, radiopharmaceutical research, materials research, workforce development, and support facilities. The project team intends NextGen MURR to become an asset of national strategic significance.

Presented by Kevin Savabi on behalf of Rubin & Hix.

Recurrent fractures of precipitation-hardened stainless-steel retaining rings were identified in the National Bureau of Standards Reactor refueling plug. Although the rings carried minimal static load relative to rated capacity, repeated failures indicated an active degradation mechanism rather than simple overload or isolated defects. The investigation combined inspection findings, finite-element stress analysis, mechanical load testing, galvanic-corrosion assessment, material characterization, hardness testing, and accelerated chloride-exposure experiments. The evidence supports stress-corrosion cracking as the most probable mechanism, driven by sustained tensile stress, a susceptible substituted alloy, suspected chloride deposits, and galvanic degradation within the bearing assembly. Mechanical testing produced localized plastic deformation without fracture, while accelerated chloride testing fractured as-installed stressed rings. The corrective redesign converted the bearing assembly from a groove-retained, corrosion- sensitive configuration into a mechanically retained Nitronic 60 bearing system, eliminating the conditions that enabled the failures.

This presentation will discuss the purpose and scope of the Department of Energy – University Fuel Services (UFS) Program. Personnel involved in the program will be introduced and contact information will be provided for team members. Information will be provided to conference attendees regarding the status of the core activities of the program. These activities include fresh fuel element fabrication and irradiated nuclear fuel shipment returns to the DOE. Current and future issues pertinent to the UFS program will also be presented. The UFS program maintains fuels support contracts and provides nuclear reactor fuel at no or low cost to 24 U.S. universities operating a total of 25 reactor facilities. These facilities include:

• Twelve Training Research Isotope General Atomics (TRIGA) facilities
• Eight plate-fueled facilities
• Three Aerojet-General Nucleonics (AGN) facilities
• One Pulstar-fueled facility
• One critical facility

The title for the fuel remains with the United States government, and, when the universities are finished with the fuel, the fuel is returned to the United States government for long-term storage. Mission of the University Fuel Services Program: The UFS Program is funded by the U.S. Department of Energy, Office of Nuclear Energy, and is managed by Idaho National Laboratory (INL) in Idaho Falls, Idaho. The program goals are:

• Keep all U.S. operating university reactor programs supplied with nuclear fuel.
• Provide assistance for movement of irradiated nuclear fuel from U.S. universities after the DOE receipt facility

authorizes the fuel receipt.

Presented by Kevin Savabi on behalf of Rubin & Hix.

The 2021 NBSR fuel failure involved a fuel element that was not securely latched, resulting in displacement, inadequate cooling, and fuel damage. Procedural improvements and visual-verification methods were introduced after the event. Further engineering review identified an opportunity to improve the mechanical robustness of the Fuel Element Head. The new Fuel Element Head, FEH V2, preserves existing fuel-element geometry, reactor interfaces, handling tools, and operating sequence. Its primary feature is a mounting-block hard stop that limits unintended back-rotation caused by spring relaxation and maintains secure engagement. The stop shifts the wear interface from the reactor structure to a controlled, replaceable component and reduces dependence on the condition of the Upper Grid Plate groove. The redesign followed a systems-engineering process including requirements, interface control, failure-mode analysis, concept evaluation, engineering analysis, detailed design, fabrication, and validation testing. Structural, thermal, hydraulic, seismic, vibration, materials, and interface evaluations confirmed adequate margins. Comparative testing showed secure engagement under all evaluated normal and off-normal handling conditions.

Research and test reactor (TRTR) facilities rely on experienced partners to execute complex projects involving spent nuclear fuel transportation, radioactive source recovery, reactor support, engineering, and specialized fabrication. This presentation describes how Secured Transportation Services (STS) integrates these disciplines into a single project delivery approach intended to reduce schedule risk, improve communication, and support mission success for domestic and international research reactor programs.

Drawing on operational experience from multiple U.S. Department of Energy (DOE) and National Nuclear Security Administration (NNSA) projects, the presentation illustrates the planning and execution of research reactor support activities, including spent nuclear fuel transportation campaigns, international radioactive source removals, route assessments, packaging and transportation planning, specialized equipment fabrication, stakeholder coordination, and regulatory interface. Common project challenges and the strategies used to address them are discussed as a means of mitigating technical, logistical, and transportation risks on future projects.

Integrated project management has improved coordination among reactor operators, regulatory authorities, transportation providers, engineering organizations, and end users. Early stakeholder engagement, comprehensive route and equipment assessments, concurrent engineering support, and proactive schedule management reduced execution uncertainty and enabled efficient completion of complex nuclear transportation and reactor support activities. The lessons learned also identify opportunities to streamline future projects through standardized planning processes, clearer communication pathways, and expanded technical capabilities.

As many research reactors operate beyond their original design life while supporting isotope production, research, education, and national security missions, demand for experienced, multidisciplinary project support continues to grow. The presentation shares practical lessons learned from complex TRTR support projects and invites discussion with the reactor community on emerging needs, collaborative opportunities, and capabilities that can strengthen the long-term success of research reactor programs.

The University of Missouri Research Reactor anti-siphon system uses compressed air that can be delivered to the primary coolant system through two redundant valves. A coupled drywell detects water within the compressed-air volume and initiates a rod run-in if water exceeds a technical-specification limit. The displacement level sensor used since 1974 has exhibited behavior that calls its reliability into question. A laser time-of-flight level sensor is being evaluated as a replacement. The sensor emits a pulse-modulated coherent light beam, starts a timer, and measures return time from the target surface. The approximately 4.9-meter target distance and 6.35-centimeter-diameter drywell require precise alignment to avoid sidewall reflections. An alignment tool inspired by a commercial tribrach was designed, and testing is planned using a mock drywell and the actual sensor.

Framatome and General Atomics established a joint venture in 1995 to produce TRIGA fuel for the United States and international markets. The production facility is located at the Framatome CERCA Category I plant in Romans-sur-Isere, France. Initial work focused on conversion of TRIGA reactors from highly enriched to low-enriched uranium. In 2013, with assistance from the U.S. Department of Energy, the facility was renovated to current standards with capacity to produce several hundred TRIGA fuel assemblies for U.S. universities. Since then, multiple U-Zr-H fuel versions have been produced and delivered. In 2026, production of the MARVEL first-core load began, with delivery planned in the same year. Under current planning assumptions, the facility will be fully loaded through 2028, with additional university-fuel options under discussion.

Objective: Demonstrate a practical method for incorporating legacy materials from the Annular Core Research Reactor into a newly manufactured regulating rod. A fabrication campaign recovered legacy reactor materials and integrated them with newly fabricated components. The effort required development and qualification of handling, closure welding, gas sampling, inspection, and acceptance methods compatible with existing geometry and materials. Component designs and fabrication steps were refined to improve manufacturability, weldability, and confidence in final assembly quality while preserving required form, fit, and function. The campaign produced a newly assembled rod containing recovered legacy materials and established a practical fabrication and inspection basis. The work demonstrated the importance of coordination among engineering, welding, nondestructive examination, metrology, and reactor operations. The resulting process basis and lessons support future fabrication development, examination planning, and irradiated-hardware refurbishment.

Objective: Explore a uranium-actinide fuel cycle for research reactors that uses zone-melt refining for cost savings, waste reduction, and nonproliferation benefits. Zone-melt refining is a nonchemical bulk-separation method that can selectively concentrate uranium, plutonium, and minor actinides from used fuel without concentrating plutonium alone. The process removes most fission products from the uranium-plutonium alloy and may substantially reduce the material volume requiring chemical reprocessing. With metallic fuels, it may eliminate chemical refining from parts of the recycling scheme. Analytical models suggest fuel-cost reductions as high as 70% and high-level-waste-volume reductions as high as 88% compared with once-through cycles. Much of the remaining waste could return to radioactivity comparable to uranium ore after about 330 years rather than more than 10,000 years. NuScale Power plans to pursue zone-melt-refining development and is interested in university partnerships for an on-campus or near-campus pilot facility associated with a microreactor deployment.

On February 3, 2026, the PSU Radiation Science and Engineering Center produced a 1 Ci Na-24 radiotracer in the form of activated sodium carbonate (Na2O3) contained in an aluminum alloy Swagelok vial. This tracer was removed from the dry tube and shipped to the customer in a Type A container. The shipping container was returned to the university on February 5th, 2026, and was discovered to be contaminated with significant quantities of Cd-115, which was not an isotope expected from the irradiation based upon the material irradiated or past experience with the procedure. The delivery vehicle and driver were recalled to the facility for a radiation survey and no contamination was identified inside the vehicle or on the driver. This event was reported to the NRC and is documented in Event Report 58155. Follow-up investigations involved testing to determine the source of the cadmium, including test irradiations and smears of the vials and dry tubes used in sodium production, NAA of smears of the tools and work surfaces used to prepare the samples, identification of issues with handling of radioactive materials and contamination control, anomalies in the transportation of the cask, and the use of x-ray fluorescence testing to interrogate the components of the Swagelok capsules. This presentation will provide a detailed summary of the event, including contributing factors, lessons learned, corrective actions, and the testing which was performed to identify the source of the cadmium contamination.

A series of mini-plate experiments has been irradiated in the Advanced Test Reactor at Idaho National Laboratory to support qualification of 19.75% enriched U-10Mo monolithic fuel for research-reactor applications. The experiments were irradiated in the I positions, B positions, and South Flux Trap to provide conditions representative of multiple fuel- plate geometries. Pre-irradiation projection analyses established safety and programmatic targets for fission density, power density, and burnup. This summary focuses on the as-run neutronic analysis following eight ATR cycles, providing validated end-of-irradiation conditions for each fueled mini-plate in support of post-irradiation examination. The analysis uses the same three-dimensional MCNP full-core ATR model used in projection analysis, updated to reflect actual operating histories. Each plate is modeled with a fuel-region mesh to resolve local burnup and fission density. MCNP-generated fluxes and replacement cross sections are used in SCALE depletion calculations, with burn steps aligned to actual cycle length, measured lobe powers, and reactor-control positions. As-run results are presented for all irradiation and power conditions. Plate-specific power density, fission density, U-235 burnup, percent FIMA, and neutron flux are reported and compared with pre-irradiation projections. The analysis shows the extent to which programmatic targets were met and provides the validated basis for interpreting post-irradiation examination measurements and supporting ongoing U-10Mo fuel qualification.

The NIST Center for Neutron Research confinement building houses the NBSR reactor and serves as a fission-product boundary to protect the public from radiological releases during an emergency. Over almost 60 years of service, the confinement-building leak rate increased. In 2025, the NCNR chose to refurbish ventilation-system automatic closure valves that seal the confinement building upon a major scram because the valves were identified as contributors to the increased leak rate. From start to finish, the project took approximately eight months. The objectives, challenges, and results of the project will be discussed along with future projects aimed at further reducing building leak rate.

Presented by Daniel LaBrier (Idaho State University) on behalf of Mary Lou Dunzik-Gougar.

High-assay low-enriched uranium (HALEU) fuels are critical to many advanced reactor designs; however, benchmark-quality data for systems fueled in the 10–20% enrichment range are relatively scarce. The Aerojet General Nucleonics Model 201-M (AGN-201M) reactor at Idaho State University is a unique HALEU-fueled training reactor (licensed for operations up to 5 W thermal power) that provides an opportunity to address this data gap. The AGN-201M’s core uses uranium dioxide fuel enriched to approximately 19.75% U-235 in a polyethylene matrix, forming a very low-power critical assembly suitable for hands-on education and research. The AGN reactor contains multiple access ports that are typically used for irradiating and characterizing material. The low power and optimal hydrogen-to-heavy element (H/X) moderation make this reactor highly sensitive to non-fissile material insertions into the experimental ports, thereby allowing reactivity measurements. A benchmark of the reactor is important for validating neutron transport and criticality simulation codes against a well-characterized thermal, graphite-reflected HALEU system, thereby increasing confidence in the modeling of next-generation reactors. The goal of this project is to develop an International Criticality Safety Benchmark Evaluation Project (ICSBEP) evaluation for the AGN-201M reactor. In Phase 1 of the project the reactor and shielding materials were disassembled for the first time in its 50+ year tenure at Idaho State. Reactor components were characterized in terms of dimensions, mass, and chemical composition. In addition, the fuel was radiographed to determine UO2 distribution within the polyethylene and was analyzed for isotopic composition. The detailed results were fed into a reactor model. Phase 2 of the project included a series of experiments with different critical configurations, including order of fuel disk placement in the core and positions of the fueled control rods. Results are being analyzed for inclusion in the benchmark report.

Presented by Ted Goodell on behalf of Andrew Allison and Dylan Ramirez.

This presentation will focus on updates on the University of Utah TRIGA nuclear reactor since TRTR 2025. These updates include the reassembly of the TRIGA reactor core following recoating the reactor tank, installation of a corrosion protection system to prevent further corrosion of the reactor tank, and approach to criticality with a new reactor core model. The presenter will also discuss observed water chemistry issues occurring during reactor operations and discussion amongst the community of suspected cause for the chemistry issue.

High-assay low-enriched uranium fuel conversions and new HALEU research-reactor designs require analysis methods that accurately capture neutronic, depletion, and thermal-hydraulic behavior. A code suite developed for research and test reactors supports plate- and rod-type fuel geometries across a wide range of powers, burnups, and flow conditions. The framework includes whole-core analysis for fuel management and burnup, coupling ADDER depletion and complex multi-cycle fuel management with MCNP neutron transport. Steady-state thermal hydraulics and hot-channel uncertainty methods use PLTEMP/ANL, including stripe-to-stripe mixing and up to three-dimensional heat conduction. STAT7 supports Monte Carlo uncertainty propagation, and PARET/ANL models core-level transients by coupling point kinetics with thermal-hydraulic feedback. The software framework has been developed and validated under NQA-1 requirements. Together, the methods provide an efficient basis for evaluating reactor performance, dentifying limiting conditions, and quantifying safety margins for HALEU conversions and new-build research-reactor applications.

The objective of this work is to examine slew-rate degradation in a circuit while key transistors are irradiated individually and measured in situ. An earlier test campaign found that the LM741 slew rate degraded with neutron fluence at the Nuclear Engineering Teaching Laboratory Beam Port 1-5 facility. Computer simulations showed that two PNP transistors in the input stage, Q3 and Q4, had the largest effect on slew-rate degradation. A discrete LM741 circuit was assembled, and the key PNP transistors were placed in sockets on a perforated board. The transistor socket was connected through 30-foot PS/2 cables to an identical board with empty transistor sockets. The first board and transistor were placed in the reactor while the second board and circuit remained outside the reactor. The long cabling and multiple socket connections significantly degraded both positive and negative slew rates. The positive slew rate was too degraded for useful analysis. The negative slew rate degraded when Q4 was irradiated and when Q3 and Q4 were irradiated together, but did not show significant degradation when Q3 was irradiated alone. The in- situ method demonstrated that Q4 irradiation affected negative slew rate. Future modifications should optimize cable length, impedance, and shielding to improve positive slew-rate measurements while keeping the external circuit outside the high-flux area. Sandia National Laboratories is managed and operated by NTESS under DOE NNSA contract DE-NA0003525.

SRAM devices are susceptible to memory errors caused by neutron radiation. Neutrons can interact with atomic nuclei in semiconductor material, generating localized charge that alters the stored state of an SRAM memory cell. Interactions with silicon can produce recoil atoms, while thermal-neutron capture by boron-10 can produce lithium and helium ions. This project seeks to quantify neutron effects on SRAM devices using a system designed to monitor memory errors in real time at the TRIGA reactor at The University of Texas at Austin Nuclear Engineering Teaching Laboratory. The device writes all-ones, all-zeros, and alternating checkerboard patterns to the SRAM chip while reactor power is increased incrementally, recording neutron-induced memory errors. The device dimensions were selected to fit within the six-inch- diameter radiation chamber.

The Nuclear Engineering Teaching Laboratory at the University of Texas at Austin has developed the capability to test analog electronics under combined neutron and RF stress via direct pin injection. Tests can be performed actively (device biased in-situ with RF injection in the reactor cavity) or passively (RF injection on previously neutron irradiated parts). This capability is being used to better understand how neutron displacement damage (nDD) modifies the response to radio frequency (RF) insult of analog microelectronics. The research literature is sparse regarding the topic of RF or electromagnetic (EM) effects on electronics exposed to ionizing radiation. In many extreme environments, such as satellites near radiation belts or nuclear reactors fielded in space or military missions, noisy EM environments are present. Ionizing radiation exposures over the mission lifetime may alter the response of sensitive electronics to EM/RF induced noise and coupling. Testing has been performed on bipolar junction transistors (BJTs) by neutron irradiating the devices to fluence levels of 5×10¹², 1×10¹³, and 3×10¹³ n/cm² 1 MeV(Si) and then later injecting RF to explore susceptibilities. The BJTs were wired in common-emitter mode, and RF was injected into the base with a bias tee. RF sweeps were performed from 10 MHz through 10 GHz and -40 – 20 dBm (2.2 mV – 2.2 V rms). Gummel measurements (base current, collector current and gain) were performed at every sweep interval. The BJTs showed significant gain compression under RF injection and exhibited enhanced gain compression for RF injection into neutron irradiated devices. This work is being performed in collaboration with Sandia National Laboratories (SNL). Future work will test diodes in a similar manner and model these effects in SNLs device simulation code, Charon.

Since 1965, Idaho State University has operated an Aerojet-General Nucleonics AGN-201 reactor, a five-watt thermal reactor used for academic and industrial research, experiments, and operator training. The reactor allows students to operate a nuclear reactor on campus and complete a rigorous training course leading to a U.S. Nuclear Regulatory Commission reactor-operator license. In early 2026, Idaho State University and NuCube Energy were accepted into the U.S. Department of Energy Reactor Pilot Program to site, develop, and operate an advanced reactor on campus. The Idaho State University Advanced Research and Testing Reactor Facility will build on NuCube's NuSun technology. The four-megawatt thermal reactor uses TRISO particle-fuel compacts, heat pipes for thermal management, and few moving parts. The presentation discusses the path toward building an advanced reactor on a state-funded university campus, the public-private partnership, lessons from authorization, and the proposed pathway to NRC licensure under 10 CFR 50.21.

There are more than 220 research reactors worldwide, with several more being constructed. A series of comprehensive PowerPoint presentations was prepared so stakeholders could use them in their own countries for research-reactor outreach programs. All presentations were developed using standard IAEA templates and placed on The University of Texas computer cloud. At the end of the project, the presentations will be uploaded to dedicated IAEA e-learning courses. The presentations developed include: History of Nuclear Science and Impact of Nobel Prize Winners; Outreach Program for Research Reactors; Strategic Outreach Plan for Nuclear Research Reactors: Products and Services Focus; Strategic Outreach Plan for Nuclear Research Reactors: Research and Development; Research Reactor Publications; European Research Reactor Conference and the National Organization of Test, Research and Training Reactors; Neutron Radiography; Medical Isotopes; Neutron Activation Analysis; and Basic Training in Support of Nuclear Energy.

The demand for precise, localized, real-time in-core neutron-flux monitoring continues to grow. Micro-Pocket Fission Detectors are miniaturized gas-filled ionization chambers with tailorable fissile coatings capable of withstanding high- temperature, high-radiation reactor environments. This work presents design optimization, testing results, and commercialization outlook for commercially configurable MPFDs. Manufacturing updates have moved the detector from laboratory prototypes toward commercial manufacturing by resolving fabrication and reliability challenges. Prototype single- and multi-chamber alpha-MPFDs using internal Am-241 sources were constructed. Baseline pulse-height spectra measured in 2023 are compared with 2026 retest data to assess long-term hermetic-seal viability, possible mineral-insulated cable outgassing, and fill-gas degradation. The presentation also discusses scalable batch manufacturing and planned deployments in university and national- laboratory test reactors. High-fidelity distributed flux mapping without significant perturbation may make the technology a cost-effective tool for irradiation experiments and next-generation reactor validation.

The deployment of advanced nuclear systems, including Small Modular Reactors (SMRs) and Advanced Modular Reactors (AMRs), requires instrumentation capable of operating reliably under increasingly demanding thermal and irradiation conditions. In this context, high-temperature fission chambers have emerged as promising neutron flux detectors for both in-core and ex-core applications. Their qualification under representative operational environments is therefore a critical step towards ensuring accurate neutron monitoring, reactor protection, and long-term system performance. This study presents an extensive characterization of high-temperature fission chambers, with particular emphasis on their behaviour at ambient temperature and under elevated temperature conditions, both under neutron irradiation. The experimental campaign was designed to assess key performance indicators, including neutron sensitivity, pulse shape and amplitude (impulse response), leakage current. Measurements were first conducted at room temperature to establish reference performance metrics and to verify intrinsic detector properties such as baseline leakage current, intrinsic noise, and nominal sensitivity. Subsequently, the chambers were subjected to high-temperature environments representative of advanced reactor operating conditions (ex-core and in-core configurations). Under irradiation, the evolution of detector response was continuously monitored as a function of temperature. The results demonstrate that the high-temperature fission chambers maintain stable sensitivity and acceptable leakage current levels up to the investigated temperature limits. Although a moderate increase in leakage current was observed with temperature, the signal-to-noise ratio remained compatible with reliable neutron flux monitoring. Pulse characteristics showed limited distortion, confirming the robustness of the detection principle under combined thermal and irradiation stress. Overall, the findings support the feasibility of deploying high-temperature fission chambers as key neutron instrumentation for SMR and AMR applications, contributing to enhanced operational safety and monitoring capabilities in next-generation nuclear reactors.

Measurements were carried out at the VR-1 research reactor of the Czech Technical University in Prague (CTU) to evaluate the operation of the Libera MONACO 3 neutron instrumentation system with two different detector types: an RJ1300 fission chamber equipped with a separate bias cable and a KNK-15 compensated fission chamber. The experimental campaign included threshold optimization, linearity measurements over a wide reactor power range, and reactor power-ramp experiments covering pulse, Campbell, and current operating modes. The measurements were performed to assess the response of the Libera MONACO 3 system under different reactor operating conditions and to compare the behavior of the two detector types when coupled to the same instrumentation platform. Particular attention was given to the transition between operating modes, detector response linearity, and the dynamic response during reactor power changes. The paper presents the experimental setup, measurement methodology, and comparative results obtained with both detector types. The measured detector responses in pulse, Campbell, and current modes are discussed, together with their applicability to reactor power monitoring across the investigated operating range.

Presented by Tim Hossain (Cerium Laboratories LLC) on behalf of Dale Julson.

The Neutron Intercepting System-on-a-Chip is a solid-state neutron-sensing platform under development for neutron detection, radiography, and dosimetry. Unlike gas-filled proportional counters or scintillation systems, NISoC integrates boron-10 neutron-conversion materials with CMOS-compatible semiconductor electronics in a compact architecture intended for high-volume manufacturing and reduced size, weight, power, and cost. Fabricated and packaged prototype devices were evaluated under thermal-neutron irradiation at The University of Texas at Austin Nuclear Engineering Teaching Laboratory. Dosimetric measurements showed repeatable response across about five orders of magnitude in thermal-neutron fluence, with a power-law dependence suitable for quantitative exposure assessment. Thermal-neutron radiography demonstrated spatial resolution below five micrometers. The architecture supports on-chip readout, low-power operation, nonvolatile retention of interaction information, and conventional electronic packaging. These results demonstrate feasibility as an integrated semiconductor platform for neutron instrumentation.

The U.S. Department of Energy Reactor Pilot Program has catalyzed advanced-reactor development. Mirion Technologies and Paragon Energy Solutions supplied much of the nuclear instrumentation for successful demonstration startups, providing a perspective on instrumentation challenges facing next-generation reactor designs. Many advanced-reactor companies rely on research reactors for material, fuel, and component testing. This presentation shows how high-quality instrumentation can support research reactors modernizing neutron-flux monitoring systems and advanced-reactor developers whose testing depends on those facilities. Upgrade pathways include neutron-flux detectors such as wide-range fission chambers and proportional counters, digital signal-processing platforms such as Mirion proTK and the DWK 260 unit, and Paragon's analog CoreVision neutron-flux monitoring system. Example measurement results from recent criticality campaigns will also be presented.

This work tests the prototype Nuclear Qualification Device developed by Idaho National Laboratory as a repeatable platform for evaluating advanced neutron sensors and quantifying uncertainty under varying thermal and radiation conditions. The prototype is a rigid, low-reactivity-worth experiment rig equipped with cable heaters designed to heat the target region to 400 C, multiple ports for neutron sensors and dosimetry packages, and thermocouple locations for heater control and temperature monitoring. Experiments were performed in both exposure rooms at the Armed Forces Radiobiology Research Institute under varying powers and temperatures. Testing included dosimetry packages and prototype self-powered neutron detectors in a thermal spectrum and a cadmium-filtered spectrum. The device maintained desired temperatures and provided live thermocouple and detector data without electronic failure. Dosimetry data characterized the neutron spectrum in each room, and detector data are being used to analyze neutron-energy-dependent response functions of different emitter materials. The first experiment demonstrates the value of the device for advanced sensor testing and supports future iterations intended to operate at temperatures up to 700 C.

The objective of the High Flux Isotope Reactor Code of Record project is to assemble a documented, authoritative set of codes, standards, requirements, and design bases governing HFIR design, operation, maintenance, and safety basis. The Code of Record is intended to be a functional list that establishes the technical basis for inclusion, identifies applicable code years, and links requirements to safety-related equipment in the HFIR Master Equipment List. Oak Ridge National Laboratory's Research Reactors Division is developing the Code of Record through structured review of authorization-basis documents, design and procurement records, modification history, regulatory requirements, and subject-matter-expert input. This paper provides an overview of planned scope, methods, and results to date.

Presented by Christine Norman (University of Missouri) on behalf of Ethan Schroeder.

The University of Missouri is developing a quality-assurance program for NextGen MURR, a proposed next-generation research reactor envisioned to operate at 20 MW thermal or greater using an open tank-in-pool configuration and high-density U3Si2 low-enriched uranium fuel. The initiative will pursue Class 104(c) licensing as a research reactor under 10 CFR 50.21 and use guidance from Regulatory Guide 2.5 and ANSI/ANS-15.8 to develop a QA program meeting 10 CFR 50.34 requirements. Members of the design and licensing team have existing QA processes, while the university's program will be implemented through multiple deployments. The initial deployment includes the Quality Assurance Program Description and procedures needed for the design and licensing phase. The completed program will draw on the expertise of the university, the design and licensing team, and the owner's representative and will be consistent with 10 CFR Part 50.34, NUREG-1537, and ANSI/ANS-15.8.

The startup-approval process for the Deployable Energy UNITY reactor was governed by DOE-STD-1271-2025 and NE Order 425.1, which define a structured, risk-aligned framework for design integration, safety analysis, and readiness verification. DOE-STD-1271 establishes early negotiation of requirements through a Nuclear Safety Design Agreement, followed by preliminary and final documented safety analyses integrating design maturation with hazard evaluation. NE Order 425.1 provides the readiness-review process for Hazard Category 1 and 2 facilities, requiring verification of system operability, safety-management-program implementation, and safety-basis adequacy before startup authorization. This framework aligns with emerging risk-informed, performance-based licensing concepts for microreactors and other low-consequence designs. UNITY's DOE authorization, grounded in negotiated requirements, integrated safety analysis, and formal readiness verification, prepared the reactor for initial operation and may provide foundational inputs for later NRC licensing.

Following a fuel-failure event in February 2021, the NIST Center for Neutron Research initiated and completed major actions including NRC regulatory activities, restart permission, three license amendments, satisfaction of a Confirmatory Order, multiple reactor-cleanup campaigns to remove loose fissionable material, refurbishment of the 15-ton reactor refueling plug, improvements to the reactor helium sweep system to reduce effluents, initiation of a Corrective Action Program and safety-culture monitoring and improvement program, and redevelopment of training and procedure-use programs. This presentation outlines completion of those actions, culminating in reactor restart to 20 MW operation in May 2026.

Nuclear reactor facilities are required to maintain detailed operational logbooks to document reactor safety checks, status, operator actions, and abnormal events. Traditionally, these records are maintained using paper-based systems, which are bulky, difficult to manage and reference, and prone to illegible handwriting. To address these limitations, we developed and tested a digital logbook at the NETL TRIGA Mark II Reactor facility. The system provides the same capabilities as the physical logbooks, as well as efficiency enhancements made possible by a digital platform. The digital log contains a centralized platform for recording startup and shutdown checks, experimental facility insertions, scrams, and operational activities while maintaining secure user authentication. It was designed to integrate with existing reactor workflows, minimizing disruption to established operating procedures while improving access to records and consistency. If paper copies are needed, the system automatically formats the digital log to replicate the physical version as a PDF. Compared with commercial electronic logbook solutions, which can require implementation costs starting at ten thousand dollars with thousands in yearly upkeep fees, this system was developed using existing institutional resources and can be implemented at no additional cost. The resulting solution eliminates software licensing fees while enabling automation of certain mandatory reporting that burdens reactor operators. This project demonstrates that a cost-effective, facility-specific digital logbook can provide the advantages of a commercial system without a significant financial investment. The approach offers a practical pathway for reactor facilities seeking to modernize operational recordkeeping, improve regulatory readiness, and use digital infrastructure to support long-term operational efficiency.

NC State has developed a hybrid-style of equipment surveillance tracking and documentation to improve information accessibility and simplify maintenance planning. The use of lightweight cloud-based tools provides a balance of automation functionality and flexibility which can work alongside existing recordkeeping systems at a facility.

Natura Resources is deploying the MSR-1 to demonstrate our liquid fueled molten salt reactor technology at Abilene Christian University. The MSR-1 is the only liquid fueled molten salt reactor permitted by the US NRC for construction. As such, many of its design features have been publicly disclosed and need not be reiterated. The overall research goals of the project are worth disclosing in detail.

The MSR-1 enables the collection of data crucial to design, operation, and licensing of Natura’s commercial MSR-100. The MSR-100 generates 100 MWe of clean, carbon-free energy. Natura is pursuing MSR-100 deployments at the RELLIS campus at Texas A&M University and in the Permian Basin, the latter deployment targeting produced water. The MSR-100 and the MSR-1 share many design features; modular design, walk-away safety, fuel salt composition, materials and construction. These features enable data collected from the MSR-1 to be directly applied to the MSR-100.

The most important unresolved question associated with liquid fueled molten salt reactors is the gaseous source term released during a rupture of the primary fuel salt bearing container. Experimental work associated with the MSRE demonstrates that the fission products aside from Xe and Kr are not present to a large degree within the gas phase. This observation is supported by analytical predictions. The MSR-1 will be equipped with instrumentation sufficient to measure the composition of the gas head space during operation. The MSR-1 also incorporates an online fuel salt sampler and coupons. This equipment enables accurate measurement of fission product behavior throughout the reactor. The MSR-1 will demonstrate redox potential control and monitoring, both in-situ and via sampling. These processes will inform the fuel salt management program associated with the MSR-100.

The MSR-1 is more than a chemistry tool, it will also test mechanical components. The MSR-1 utilizes mechanical fuel salt valves and pumps. The MSR-1 also includes an off-gas system and fuel purification system. The performance history of this equipment will inform the MSR-100.

This work presents the develop of a digital twin framework (denoted as Keeper) for demonstrating autonomous operations the Idaho National Laboratories Neutron Radiography (NRAD) reactor. Keeper is built to incorporate diagnostics, prognostics, and control within a digital twin framework to provide a system-level awareness of NRAD capable of facilitating autonomous operations. The goal of this work is twofold; develop and demonstrate autonomous operations on a reactor from startup, normal operations, off-normal operations, and shutdown along with the ability to document the process for achieving Department of Energy regulatory compliance. For achieving autonomous operations, a combination of physics- and machine learning-based algorithms provide a baseline for comparing the expected state of the reactor and predicting the future state of the reactor. Once the current and future state of the reactor are established, a combination of traditional control (i.e. proportional-derivative controllers), model predictive control, and reinforcement learning algorithms are used to determine how controls rods should move to achieve a set overarching reactor operator determined goals (reactor power, minimize control rod movements, etc.). To allow signals into the NRAD core, documentation of Keeper and its implementation was capturing via a Concept of Operations and Test Plan. These documents, along with a series of shorter documents were used to show compliance with DOE-STD-1271-2025 which dictates how autonomous operations at DOE-licensed facilities can be performed.

This effort seeks to leverage current work being performed for digital twins at research reactors across the university and national laboratory system. Multiple universities and laboratories have developed frameworks, algorithms and approaches for digital twins independently leading to a wealth of knowledge which tends to only be shared in a disparate nature. This often leads to multiple entities running into similar roadblocks that others may have already accomplished. To accelerate the development of digital twins across research reactors, we are seeking to develop a working group which can collaborate, build off each other’s successes, and overcome larger obstacles than we could individually. Research reactors represent a uniquely valuable, and largely untapped, proving ground for digital twin technology providing a lower-consequence, higher-access environments than industry power reactors, making them an ideal test bed for iterating on digital twin methods before they mature toward deployment in commercial systems.

The proposed consortium would provide a shared platform for this maturation. Rather than duplicating infrastructure at each site, member institutions could contribute to and draw from common resources, including:

• Shared benchmarks and test cases — a common set of operational scenarios, transients, and datasets against which different digital twin approaches can be validated and compared.
• Open frameworks and tooling — reusable software components for data ingestion, model reduction, state estimation, and visualization.
• A community repository of lessons learned — documenting successes and failed approaches/dead ends to avoid repeating mistakes.
• Cross-institutional verification and validation practices — working toward shared standards for what constitutes a validated digital twin in the research reactor context.
• Workforce development — giving students and early-career researchers exposure to multiple reactor types, instrumentation schemes, and digital twin architectures.
• Shared understanding of value statements — Digital twins are being used to improve software accuracy, qualify software for applications, improve operational and experimental efficiency, and enhance education; this will help identify and realize more value for each institution.

By coordinating rather than duplicating efforts, the consortium aims to compress the timeline for digital twin development at academic and research institutions allowing for more meaningful collaboration and research into digital twin capabilities.

The radiation does rate mapping system, RAD-MAP, was developed as a student capstone project along with support from the McMaster Nuclear Reactor (MNR) to survey areas around the reactor core in preparation for a new beam port project. The system accurately measures and maps the position of the submersible and the measured dose rate on to a live 3D virtual map that can be clearly read by an operator. RAD-MAP consists of two underwater cameras that as fastened to support structures, a submersible remote operated vehicle (ROV) with a traceable marker and mounted dose rate probe, an AprilTag reference marker the for the visual triangulation system, and a python script that tracks the ROV position, records dose rate measurements, and executes data mapping. During a survey, an operator controls the ROV within the reactor pool while position and dose rate data are automatically recorded, mapped, and displayed in live time. This system provides a visual representation of potential radiation hazards present in the proposed work are, supporting and improving survey planning and the development of a safer work environment.