DNCSH

The Seven Percent Critical Experiment (7uPCX) fuel inside the Sandia Pulsed Reactor Facility for Critical Experiments (SPRF/CX) core tank. The core tank is filled with water to conduct water-moderated critical benchmark experiments. The new HALEU UO2 fuel being procured for this project will have the same dimensional requirements at the 7uPCX fuel.

Sandia National Laboratories (SNL) proposes new critical benchmark experiments to expand on the existing experiments performed at the Sandia Pulsed Reactor Facility for Critical Experiments (SPRF/CX). These experiments will require the use of UO2 fuel rods that meet the same dimensional requirements of the current Seven Percent Critical Experiment (7uPCX) fuel rods but with increased U-235 enrichment to nearly 20 wt.% (compliant with the existing Safety Basis for SPRF/CX). 

This will target the need for high quality integral critical benchmark experiments within the enrichment range of 10-20 wt.% U-235. The proposed design allows for reuse of much of the equipment and non-fissionable materials (e.g., grid plates, safety/control rods, support materials, molybdenum sleeves, titanium/tantalum rods) from the previous 7uPCX experiments, as well as future exploration of nuclear data for advanced moderators (e.g., using fuel arrays with interstitial BeO or graphite rods/sleeves) and non-fissionable material validation. The goal of this project is to procure new HALEU UO2 fuel rods for critical benchmark experiments at SPRF/CX. The procurement effort is a collaboration with Idaho National Laboratory (INL), where INL’s Materials & Fuels Complex will fabricate 19.75 wt.% enriched UO2 pellets to be loaded into aluminum-clad fuel rods at SNL matching 7uPCX dimensional requirements. 

Project PIs with TerraPower staff in front of the assembly during the execution of the copper reflected critical configuration.

Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL) propose a series of three new critical experiments using chlorine absorbers to address the criticality safety needs of TerraPower for molten salt fuel fabrication and transport. These experiments are an extension of a Nuclear Criticality Safety Program (NCSP) funded experiment, the Thermal / Epithermal eXperiment (TEX) with Chlorine, that is scheduled to be completed at the National Criticality Experiments Research Center (NCERC) in FY24. Specifically, these experiments will use high-density cast chloride salt absorbers in a highly enriched uranium assembly to target the thermal 35Cl(n,ɣ) absorption and fast 35Cl(n,p) absorption, which are nuclear data needs identified for the TerraPower’s Molten Chloride Reactor Experiment (MCRE) and the Molten Chloride Fast Reactor (MCFR). Current chlorine benchmarks are insufficiently sensitive to perform adequate criticality safety analyses for fuel fabrication, staging/loading, and transportation and, while not specified in this call, provide validation for power production and spent fuel transportation, storage, and disposition.

Milestones and Deliverables:

FY25: Experiment Design and Procurements

• LLNL completes and distributes to the working group the TEX-Chlorine for TerraPower (TEX4TerraPower) experimental design report
  • Completed. OSTI: https://www.osti.gov/biblio/2536828
• LLNL/LANL/TerraPower initiate procurement of additional components for experimental campaign.
  • Completed.

FY26: Inspection, Execution, and Analysis of Experiments

• LLNL completes an experimental execution report following the completion of the experimental campaign.
  • In Progress. (By June 30, 2026)
• LLNL/LANL complete inspection of new components necessary for experiment execution.
  • Completed.
• •LLNL begins the benchmark evaluation of the experimental configurations.
  • In Progress. (By September 30, 2026)

FY27: Evaluation of the Experiments for the ICSBEP

• LLNL completes the benchmark evaluations, has the benchmark reviewed, sent to the ICSBEP TRG, and resolves any comments.
  • To Be Performed.

The MARVEL Core Design

The Microreactor Applications Research Validation and Evaluation (MARVEL) at Idaho National Laboratory (INL) is a first of its kind reactor under development since 2020. It will be fueled by 19.75wt% enrichment uranium in a zirconium hydride matrix (U-ZrH) with Be/BeO reflectors and use naturally-circulating eutectic NaK as its primary coolant. An intermediate step prior to full operations is to perform a dry initial criticality. The reactor will be taken critical a subset of fuel sections prior to loading with the NaK coolant. Key system components will be present, including the reflectors, and the control drums. A benchmark evaluation of the dry criticality will be completed for inclusion in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook.

Milestones and Deliverables:

FY26: Coordination

• The INL and Lawrence Livermore National Laboratory team members will coordinate prior to the MARVEL dry criticality: 

In progress.

FY27: Evaluation

• The MARVEL dry criticality will occur in early FY27. Team members will be present to observe and gather data. A sensitivity analysis and benchmark model will be developed and reviewed internally at the INL: 

Not Started.

FY28: ICSBEP evaluation #1 re-submission and ICSBEP guidance report

• The benchmark evaluation will be provided to Argonne National Laboratory for external review. The completed benchmark will be presented to the ICSBEP TRG at the 2028 meeting. Comment resolution will follow in preparation for submittal for publication: 
By August 1, 2028

Graphite is an important material in nuclear energy systems because of its excellent neutronic properties, high-temperature stability, and ability to serve as both a moderator and structural component. It has been widely used in gas-cooled reactors and remains a key material for advanced reactor concepts, including high-temperature gas-cooled reactors, molten salt reactors, and microreactors. Because these systems operate at elevated temperatures, accurate high-temperature thermal neutron scattering cross sections are essential for predicting neutron moderation, reactor performance, criticality safety margins, and material behavior throughout reactor operation. 

Measured double-differential thermal neutron scattering section of graphite at 300 K, obtained using 30 meV incident neutrons.

However, there is currently no high-resolution experimental validation of high-temperature thermal neutron scattering cross sections in graphite. Existing thermal neutron scattering law Evaluated Nuclear Data Files (ENDF) extend to 2000 K but were generated using phonon densities of states calculated at 0 K with density functional theory or at room temperature using classical molecular dynamics. Consequently, the effects of high temperature (anharmonicity) on graphite atomic vibrations—and therefore on thermal neutron scattering cross sections—have not been adequately represented. 

This project addresses this gap by combining high-resolution inelastic neutron scattering measurements at the Spallation Neutron Source at Oak Ridge National Laboratory, high-temperature X-ray diffraction, first-principles calculations, and machine-learning atomistic simulations to determine the temperature-dependent phonon density of states of graphite from 300 to 1200 K. These measurements provide the experimental foundation for validating and improving high-temperature thermal scattering law evaluations and developing more accurate ENDF thermal scattering law data. The project provides the first comprehensive high-resolution experimental benchmark for graphite at elevated temperatures, improving confidence in nuclear data used for reactor physics, criticality safety, and licensing calculations for next-generation nuclear systems.

Milestones and Deliverables

FY25/26: High-temperature measurements and atomistic modeling

• Perform and analyze high-temperature X-ray diffraction and inelastic neutron scattering measurements to determine graphite thermal expansion, the thermal neutron scattering function, and the temperature-dependent phonon density of states.

Status: Completed.

• Perform temperature-dependent self-consistent phonon (SCPH) calculations and machine-learning force-field molecular dynamics simulations to model graphite vibrational properties and anharmonic effects.

Status: Completed.

FY26: Thermal scattering law evaluation

• Develop temperature-dependent graphite thermal neutron scattering law (TSL) evaluations and validate the new evaluations using benchmark reactor physics and criticality safety calculations.

Status: In progress.

Deliverables

• Submit the measured high-temperature double-differential neutron scattering cross sections to the Experimental Nuclear Reaction Data (EXFOR) database at the International Atomic Energy Agency (IAEA).

Status: Planned.

• Publish the high-temperature graphite vibrational properties, including experimental measurements and atomistic modeling, in a peer-reviewed journal.

Status: Manuscript completed and being prepared for submission through the ORNL RESolution system.

• Publish the temperature-dependent graphite thermal neutron scattering law evaluations and their validation in benchmark calculations in a peer-reviewed journal.

Status: Planned.

The Los Alamos critical experiments team at the National Criticality Experiments Research Center (NCERC) and the fuel fabrication experts at the Low Enriched Fuel Fabrication Facility (LEFFF) have identified means of validating fuel processing behavior which could lead to higher throughput when techniques are transitioned from LEFFF to commercial scales. Careful Analysis of Criticality in TRISO Utilizing Systems (CACTUS) will provide validation data supporting the areas of graphite and silicon carbide (SiC) and has the potential to reduce margin in future facilities. CACTUS will use uranium metal fuel; polyethylene reflectors; and graphite and SiC interstitial moderators to understand the effects of these materials separately and together.

The CACTUS critical experiment campaign will result in a new series of benchmark experiments for inclusion into the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook. CACTUS will combine highly characterized graphite, SiC, and polyethylene with existing benchmark quality metal fuel in various configurations to examine different points along the fabrication pipeline of TRISO particles. By taking a step back from fully formed TRISO particles and compacts, cleaner validation data can be obtained for each step in the fabrication pipeline. This data will directly contribute to the front-end of TRISO production and could lead to higher throughput when techniques are transitioned from LEFFF to commercial scales.

Highly enriched uranium (HEU) pancake plates and the natural uranium (NU) pancake plates used in several previous ICSBEP evaluations will be combined to create the desired high-assay low enriched uranium (HALEU) enrichment level. This level can be tuned from 10 – 19.9 wt.% overall enrichment based on the plates used. CACTUS will utilize new interstitial, moderating components of graphite and SiC that will be procured along with new polyethylene reflectors.

Milestones: 

• FY26: Final Design Report
  • In progress for September 2026
• FY27: Complete Procurement
• FY27: Complete Experiment Campaign
• FY28: Submit benchmark evaluation to ICSBEP

UML Research Reactor beam ports. The fast neutron beam port is the area where this project will be executed.

Newly developed broomstick-like experiments will be performed to measure neutron (and gamma) induced energy-dependent integral transmission data by using the neutron (and gamma) pencil beam of the fast neutron beam port at 1MW water pool UMass Lowell Research Reactor (UMLRR) that uses LEU (19.75wt% ²³⁵U with fuel type U₃Si₂ with Al cladding). The focus will be on fluorine measured as Teflon with different thicknesses. The reactor is used for a wide range of multidisciplinary educational and research purposes, including radiation effects on electronics, neutron activation analysis, neutron radiography, neutron and gamma irradiation of biological media, neutron irradiation of borated polymers, and production of short-lived isotopes for educational and research purposes, educational coursework, and personnel training. The picture of the reactor is shown below. The UMLRR facility consists of a fast beam port that is ideal for developing neutron (and gamma) transmission integral measurements in the mixed radiation field generated by the UMLRR. The fast port bunker is currently available and provides ~4×10^6 n/cm²/s of neutrons from thermal to 20 MeV (U spectrum) in a 6’’ diameter and an 81’’ long beam pipe. The beam will be collimated into a pencil beam with options of 1’’, 1.5’’ and 2’’ diameter.

Milestones and Deliverables

• Year 2026
  • [Q4]
    • Fast Port: Hardware, Supplies and Materials for broomstick measurement selected and ordered. Order gamma-ray standards for activation foils HPGe counting measurements
    • VdG: Obtain and prepare LiF and Li target material, verify stable VdG operation over full energy range, design detector mounting structure
    • Work on UMass reactor design MCNP to SCALE.
• Year 2027
  • [Q1]
    • Fast Port: The first version of the collimator installed and tested for transmission using proton recoil detectors
    • VdG: Assemble and setup detector mounting structure. Initial testing of proton-recoil and stilbene detectors with 1-MeV neutrons. Analyze and assess setup to optimize energy characterization measurements
    • Modeling and simulation with SCALE/MCNP
  • [Q2]
    • Fast Port: Foil activation measurements simulated, selected foils activated and measured, result evaluated.
    •  VdG: Perform 1st phase energy response characterization of proton-recoil detectors by directly scanning VdG accelerator beam energy.
    • VdG: Perform 1st phase energy response characterization of proton-recoil detectors by varying detector angle and using neutron angular distribution
    • Modeling and simulation with SCALE/MCNP
  • [Q3]
    • Fast Port: Final design of the collimator – identification of parts that need improvements
    • VdG: Perform 1st phase energy response characterization of stilbene detectors by directly scanning VdG accelerator beam energy.
    •  VdG: Perform 1st phase energy response characterization of stilbene
    • Deliverable 1 – Progress report on Collimator design, and detector characterization
    • Modeling and simulation with SCALE/MCNP
  • [Q4]
    • Fast Port: Changes to the first version of the collimator – identification of parts that need improvements. Procurement and testing/validation.
    •  VdG: Re-evaluate response of detectors at 1-MeV neutron energy following measurements at UML RR.
    • VdG: Optimize energy response characterization measurements based on Year 1 results
    • Modeling and simulation with SCALE/MCNP
• Year 2028
  • [Q1]
    • Fast port: production data measurements. Fluorine (Teflon) + backgrounds + foils activation.
    • VdG: Perform 2nd phase energy response characterization of proton-recoil detectors by directly scanning VdG accelerator beam energy.
    •  VdG: Perform 2nd phase energy response characterization of proton-recoil detectors by varying detector angle and using neutron angular distribution
    • Modeling and simulation with SCALE/MCNP
  • [Q2]
    • Fast port: production data measurements. Fluorine (Teflon) + backgrounds + foils activation.
    •  VdG: Perform 2nd phase energy response characterization of stilbene detectors by directly scanning VdG accelerator beam energy.
    •  VdG: Perform 2nd phase energy response characterization of stilbene detectors by varying detector angle and using neutron angular distribution
    • Modeling and simulation with SCALE/MCNP
  • [Q3]
    • Fast port: production data measurements. Fluorine (Teflon) + backgrounds + foils activation.
    • VdG: Initial transmission measurement testing using VdG accelerator-based neutrons to compare to UML RR results
    • Modeling and simulation with SCALE/MCNP
  • [Q4] 
    • VdG: Perform broomstick measurements with stilbene detectors at VdG
    • Deliverable 2 – Report on neutron beam characterization at the collimator at fast beam port of UMLRR 
    • including validation of foils activation measurements. Preliminary results on transmission measurements with CaF2
    • Modeling and simulation with SCALE/MCNP 
• Year 2029
  • [Q1]
    • Fast port: Results on transmission measurements for Fluorine submitted for publication and evaluation to ORNL.
  • [Q2]
    • Analyze and compare VdG results to UML RR results.
  • [Q3]
    • Final paper on measurements and evaluations submitted to peer review journal

The Royal Military College of Canada SLOWPOKE-2 core.

Existing data on criticality measurements taken at the Safe LOW Power (K) critical Experiment (SLOWPOKE-2) reactor at the Royal Military College of Canada will be evaluated for a benchmark to be included in the ICSBEP handbook. In 2021, the entire reactor core was removed, replaced with a core of fresh HALEU (19.75 wt%) uranium dioxide fuel pins, and taken to criticality at very low power.

Milestones and Deliverables

FY26: ICSBEP evaluation progress report

• ORNL and CNL provide a draft description of the measurements being evaluated and plan to address outstanding experimental uncertainties, to inform the decision on suitability for inclusion as a criticality benchmark. Deliverable is a technical report. In progress.

FY27: ICSBEP evaluation progress 

• The ICSBEP evaluation draft is finalized and the internal review is completed: N/A

FY28: ICSBEP evaluation submission

• Submission of the evaluation to the 2028 ICSBEP Technical Review Group (TRG) meeting: N/A

• Defend the evaluation at the ICSBEP TRG meeting in April 2028: N/A 

The Canadian Nuclear Laboratories ZED-2 reactor.

New criticality measurements with large samples of non-fissile structural materials (e.g. Ni, Ta, Mo, Fe) will be performed in the Zero Energy Deuterium (ZED-2) Reactor in Chalk River to support I.C.3. “Non-fissile material validation” to be submitted for inclusion in the ICSBEP handbook. These critical experiments will not be performed with HALEU fuel, but instead rely on fuels already available in the ZED-2 inventory, such as those employed for the measurements listed in ICSBEP evaluation LEU-MET-THERM-003 or IRPhEP evaluation ZED2-HWREXP-001. The absorbers will be exposed to a thermal neutron spectrum and their sensitivity will be maximized during the experimental design phase to increase their use for validation.

Milestones and Deliverables

FY26: Critical Experiment Design report

•  From procurement and simulation assessments estimating expected sensitivities and experimental uncertainties, ORNL and CNL provide final critical experiment design in a technical report with all information gathered and preliminary results of computer simulations. In progress.

FY27: Critical experiments are completed and ICSBEP evaluation progress 

• The experiments are completed, and a report is created by ORNL and CNL, summarizing the different critical configurations studied and the basic characteristics of the reactor and absorber elements used: N/A

• The ICSBEP evaluation draft is finalized and the internal review is completed. N/A

N/A FY28: ICSBEP evaluation submission

• Submission of the evaluation to the 2028 ICSBEP Technical Review Group (TRG) meeting: N/A

• Defend the evaluation at the ICSBEP TRG meeting in April 2028: N/A 

The Missouri S&T Reactor operating at 180 kW.

The Missouri S&T Reactor (MSTR) is a light-water moderated, pool-type research reactor fueled with U3Si2/Al fuel with a nominal enrichment of 19.75 wt% 235U. The reactor can be operated in a fully light-water reflected mode, or adjacent to a graphite thermal column. The MSTR has not yet been evaluated in either the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook or the International Reactor Physics Evaluation Project (IRPhE) Handbook, thus providing a new, uncorrelated HALEU assembly for use in validation. Evaluation of MSTR critical configurations responds to proposal topics I.C.2 and I.C.5 by filling the gap of 10-20 wt% enrichment and providing new graphite validation data.

Milestones and Deliverables

FY25: ICSBEP evaluation #1 work

• ORNL and MS&T conduct the critical operation of the MSTR in standard and thermal column (graphite) configurations for the evaluation #1, and a report summarizing the operation is released. Done. 

Related publication at PHYSOR 2026 conference: James Daniell, Ayodeji Alajo, Ethan Taber, Mathieu Dupont, and Andrew Holcomb, “Least Squares Based Filtering Method for Missouri University of Science & Technology Reactor Criticality Assessment,” Proceedings of the PHYSOR 2026 conference, Turin, Italy (2026).

FY26: Progress on ICSBEP evaluation #1 and design/planning of evaluation #2

• Evaluation#1 draft is finalized and the internal review is performed before submission to the independent reviewer in November 2026: In progress.

• Design of additional critical experiments for Evaluation #2 (E#2) summarized in a report: In progress.

FY27: ICSBEP evaluation #1 submission and evaluation #2 progress

• Submission of E#1 to the 2027 ICSBEP Technical Review Group (TRG) meeting: N/A

• Defend E#1 at the ICSBEP TRG meeting in April 2027: N/A 

• Critical experiments for E#2 are performed, and experimental data is obtained and summarized in a report: N/A 

FY28: ICSBEP evaluation #2 submission

• Submission of E#2 to the 2028 ICSBEP Technical Review Group (TRG) meeting: N/A

• Defend E#2 at the ICSBEP TRG meeting in April 2028: N/A 

The Reactor Critical Facility reactor core.

The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) is a pool-type reactor fueled with SPERT-type 4.807 wt% enriched UO2, stainless-steel clad fuel rods. The RCF is highly flexible in the allowable arrangement of fuel and in the addition of experimental components to the core. The RCF was used to perform critical experiments involving a stainless-steel pipe placed at the center of the core between 2016 and 2018. The proposed work is to create two evaluations of the basic core configuration as well as those nonfissile stainless-steel focused experiments for submission and publication in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. Many important structural materials are inadequately represented in the ICSBEP handbook, including the major components of steels. These evaluations would provide a new uncorrelated LEU assembly for use in validation of non-fission materials. Evaluations of the RCF critical configurations responds to the EAW#1 topic area I.C.3 by providing new validation data with sensitivity to different non-fissile stainless-steel elements.

Milestones and Deliverables

FY25: Progress on ICSBEP Evaluation #1 (E#1)

• Evaluation#1 draft is finalized and the internal review is performed before submission to the independent reviewer in November 2026: Done. (not on OSTI)

Related publication at NCSD 2025 conference:  M. N. Dupont, M. Eklund, and W. Ji, “Reactor Critical Experiments at the Rensselaer Polytechnic Institute Reactor Critical Facility with Noteworthy Non-fissile Stainless Steel Elements Sensitivity,” Proceedings of the NCSD 2025 conference, San Antonio, Texas (2025).

FY26: ICSBEP evaluation #1 submission and evaluation #2 progress

• Submission of E#1 to the 2027 ICSBEP Technical Review Group (TRG) meeting: Done. The evaluation was assigned to the LEU-COMP-THERM-114 label by NEA. (not on OSTI)

• Defend E#1 at the ICSBEP TRG meeting in April 2026: Done. The evaluation was not accepted for publication in the 2026 ICSBEP handbook and will be re-submitted in 2027.

• Evaluation#2 (E#1) draft is finalized and the internal review is performed before submission to the independent reviewer in November 2026: In progress.

FY27: ICSBEP evaluation #1 and evaluation #2 submissions

• Submission of E#1 to the 2027 ICSBEP Technical Review Group (TRG) meeting: N/A

• Defend E#1 at the ICSBEP TRG meeting in April 2027: N/A 

• Submission of E#2 to the 2027 ICSBEP Technical Review Group (TRG) meeting: N/A

• Defend E#2 at the ICSBEP TRG meeting in April 2027: N/A 

Enrichments in the 5-10 wt% U-235 range are currently being considered for future light water reactor designs that include the presence of gadolinium. At these enrichments, there are no benchmarks containing gadolinium. Taking credit for gadolinium in storage and transportation is a common practice and can provide additional margin during moderated accident scenarios. This proposal intends to fill this gap by providing benchmark data for these moderated systems to be used to assess the criticality of fuel during storage and transportation. 
Pacific Northwest National Laboratory (PNNL) and Sandia National Laboratory (SNL) will design a series of critical experiments to be performed at SPRF/CX that address benchmarking gaps for gadolinium. The experiments will provide data in systems containing fuel composed of 5-10 wt% U-235 that can be used to reduce uncertainty and biases when assessing the criticality of fuel in storage and transportation in which gadolinium is present. This will be accomplished using the Seven Percent Critical Experiment (7uPCX) fuel available at SPRF/CX.

Milestones and Deliverables:

Task 1: Design (FY26-FY27)

• FY26: Design Development Draft – PNNL submits draft to SNL for review.
• FY27: Design Development Final – PNNL submits final design to SNL for experiment plan development.

Task 2: Procurement and Experimentation (FY27-FY28)  

• FY27: Procurement of Gadolinium – PNNL will procure gadolinium to be provided to SRNL for fabrication of absorber rods. SNL will procure the remaining materials to fabricate the gadolinium rods. 

Task 3: Generation and documentation of ICSBEP evaluation (FY28)

• FY28: ICSBEP Evaluation – PNNL and SNL will develop the ICSBEP evaluation based on the measured critical configurations and submit the evaluation to the ICSBEP. 
   

Rendering of critical experiment configuration with carbon steel and LEU+ fuel rods.

Storage and transportation scenarios require consideration of a variety of moderation conditions. Particularly for dry fresh fuel storage arrays, a consideration is that the maximum reactivity may occur at low water densities, which may be experienced during fire-fighting or other off-nominal scenarios. There are currently no experiments designed to target these conditions. The validation of optimum moderation conditions will become important as enrichments increase, and criticality margins decrease and will fill a crucial data gap. 

Milestones and Deliverables:

FY25: Feasibility evaluation of obtaining configurations with high similarity to optimum moderation conditions with the SPRF/CX reactor

• Feasibility assessment showed that configurations with adequate similarity to optimum moderation configurations could be obtained by adding iron bearing rods to SPRF/CX configurations with the 7uP/CX fuel.

FY26: Final selection of experimental configurations and experimental Execution

• Final selection of experiments completed in March 2026
• Experiments expected to be performed in late summer of 2026 at Sandia National Laboratories

FY27: Submission of Experimental evaluation to ICSBEP Technical Review Group

• PNNL to submit the ICSBEP evaluation to TRG

Radial Cross-section of Initial Fuel Loading for THETA Configuration 3 which had both polyethylene (white) and borated polyethylene (green) interstitials.

The transportation of large quantities of high assay low enriched uranium (HALEU) TRi-structural ISOtropic (TRISO) fuel is a large bottle neck to efficient scaling of commercial nuclear power. This problem was noted by industry partner, Kairos Power, LLC, who is focused on developing and delivering Fluoride-salt cooled high temperature reactors, using HALEU TRISO fuel, to the US electric power market. To effectively scale transportation, a validation basis is needed, especially for the consideration of accident scenarios. Typically, validation for criticality safety is heavily influenced by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook, but at this time, it only includes a few benchmark series with TRISO fuel. Thus, the TRISO-form HALEU-fueled Experiment for Transport Applications (THETA) project was conceived to address the validation gaps for TRISO fuel transportation upset scenarios, particularly water infiltration of pebble fuel containers/ canisters. These scenarios have a complex neutron flux and are expected to have a bias not quantifiable by other fuel forms. Additionally, the subsequent ICSBEP evaluation to address transportation upset scenarios also improves the benchmark coverage of water infiltration upsets in fuel storage and handling applications. THETA is built on the Deimos advanced reactor testbed platform adding stainless steel reflectors to simulate the transport container, and  interstitial polyethylene and borated polyethylene to neutronically simulate water ingress into the container. The THETA experiment campaign was performed on the Comet assembly at the National Criticality Experiments Research Center (NCERC) managed by Los Alamos National Laboratory.

Milestones & Deliverables

FY25: Final Design and Experiment Execution

• Complete experiment final design and initiate procurement
  • Done
• Execute Experiment at NCERC and perform all measurements required for an ICSBEP benchmark
  • Done
  • DOE-NE press release
• Complete draft of ICSBEP benchmark section 1, detailing all knowledge of the experiment and the components
  • Done. Section 1 is near final based on feedback from Deimos benchmark at ICSBEP TRG

FY26: Complete draft of ICSBEP benchmark

• Complete ICSBEP benchmark draft report
  • On schedule for completion. Many model/ geometry features and uncertainties agreed upon with ICSBEP TRG during Deimos Configuration 1 ICSBEP TRG review.

Generalized phonon density of states (DOS) of SS316 at 35 °C, measured on the ARCS spectrometer at the Spallation Neutron Source and calculated from a machine-learned interatomic potential trained in VASP and run in LAMMPS. The calculation is preliminary.

Structural alloys are not moderators, but they scatter and reflect neutrons. Thick steel returns neutrons to fissile regions and slightly slows them, which raises reactivity. Dry storage canisters and transport casks for spent nuclear fuel are built from austenitic stainless steel, and criticality benchmarks with metal reflectors cannot be modeled correctly without the scattering physics of the reflector. Current nuclear data libraries carry no thermal scattering law (TSL, also written S(alpha, beta)) evaluations for these alloys.

This work measures that scattering and builds the missing evaluations for nine alloys: SS304L, SS316L, Inconel 617 and 625, Hastelloy X, and FeCrAl. Inelastic neutron scattering (INS) measurements on the ARCS and VISION spectrometers at the Spallation Neutron Source (SNS) give the double differential scattering cross section, and from it the phonon density of states (DOS). Transmission measurements at the VENUS imaging beamline at SNS give the total cross section.

Alloys are harder to model than ordered crystals, because the alloying elements sit at random on the host lattice and the calculated phonon spectrum is sensitive to the atomic configuration assumed. Special quasirandom structures supply the starting configurations. Total scattering measurements on the NOMAD diffractometer at SNS constrain the local structure through the pair distribution function (PDF), and machine-learned interatomic potentials trained in VASP produce the phonon spectra in LAMMPS.

The work also asks whether every alloy needs its own TSL evaluation, or whether one surrogate evaluation can stand in for the families of similar alloys. The evaluations will be tested against criticality benchmarks from the International Criticality Safety Benchmark Evaluation Project to measure how much alloy composition matters. Measured cross sections will be reported to the Experimental Nuclear Reaction Data (EXFOR) database, and the new evaluations to the National Nuclear Data Center (NNDC) for inclusion in a future ENDF release.

Milestones and Deliverables

The work spans FY26 and FY27 and finishes at the end of FY27.

FY26/F27: Measurement campaign and initial evaluations

• Sample procurement: Complete for the first six alloys.

• Room-temperature inelastic neutron scattering measurements: Complete. Double differential scattering cross sections were measured for SS304L, SS316L, Inconel 617, Inconel 625, Hastelloy X, and FeCrAl on the ARCS spectrometer at SNS.

• Low-temperature inelastic neutron scattering measurements: Complete. The same six alloys were measured at 5 K on the VISION spectrometer at SNS.

• Total scattering measurements: Complete. Pair distribution functions of the same six alloys were measured on the NOMAD diffractometer at SNS, and are used to constrain the local structure models behind the phonon calculations.

• Transmission measurements: Planned.

• Machine-learned interatomic potentials and first TSL evaluations: In progress.

• Surrogate TSL study against criticality benchmarks: Planned.

Deliverables

• Submission of measured cross sections to the Experimental Nuclear Reaction Data (EXFOR) database: Planned.

• Submission of the TSL evaluations to the National Nuclear Data Center (NNDC) for inclusion in a future ENDF release: Planned.

• Publication presenting the measurements, the evaluations, and the surrogate TSL study: Planned.

Generalized phonon density of states (DOS) of silicon carbide at 308 K, 823 K, and 1023 K, calculated from a machine-learned interatomic potential (left) and measured on the ARCS spectrometer at the Spallation Neutron Source (right). The dashed line marks the 308 K peak position.

The reactor designs supported under the Advanced Reactor Demonstration Program (ARDP) operate at high temperature, and they rely on moderator materials whose thermal neutron data were evaluated without validation data at those temperatures. Silicon carbide (SiC), zirconium carbide (ZrC), magnesium oxide (MgO), beryllium oxide (BeO), and beryllium carbide (Be₂C) appear in tristructural isotropic (TRISO) fuel particle coatings and in two-phase composite moderator concepts. The thermal scattering law (TSL, also written S(alpha, beta)) evaluations for these materials in current nuclear data libraries were built from phonon calculations performed at 0 K or at room temperature. 

This work measures the two quantities those evaluations have to reproduce, at temperature. Inelastic neutron scattering (INS) measurements on the VISION and ARCS spectrometers at the Spallation Neutron Source (SNS) give the double differential scattering cross section, and from it the temperature-dependent phonon density of states (DOS). Transmission measurements at the VENUS imaging beamline at SNS give the total cross section. Machine-learned interatomic potentials trained on first-principles data are then used for the phonon calculations at the measured temperatures, and the TSL evaluations built from those calculations are checked against both measurements.

Two capabilities developed as part of this work make the approach repeatable for other moderators. IRMA is an open-source code (https://github.com/ramic-k/IRMA) that turns one phonon calculation into ENDF-6 evaluated data, instrument-resolved neutron spectra, and scattering kernels for direct use in Monte Carlo neutron transport codes. IRMA also lowers the expertise needed to produce an exploratory TSL evaluation for a material that has none. Its pretrained machine-learned interatomic potentials produce a complete evaluation from an atomic structure file alone. VENUS is now a validated thermal neutron transmission instrument, covering 1.35 meV to 30 eV with characterized backgrounds.

Measured cross sections will be reported to the Experimental Nuclear Reaction Data (EXFOR) database, and the new evaluations to the National Nuclear Data Center (NNDC).

Milestones and Deliverables:

The work spans FY25 and FY26 and finishes at the end of FY26.

FY25/26: Measurements and evaluation work 

• Perform INS measurements: Complete. Temperature-dependent double differential scattering cross sections were measured for SiC, ZrC, MgO, BeO, and Be₂C on the VISION and ARCS spectrometers at SNS.

• Perform transmission measurements: Complete. Total cross sections were measured at the VENUS imaging beamline at SNS.

• Perform TSL evaluations: Complete. Machine-learned interatomic potentials were trained for each material, and temperature-dependent TSL evaluations were generated from the resulting phonon calculations.

Total cross section of vanadium measured at the VENUS imaging beamline at the Spallation Neutron Source (points), compared with a calculation using NCrystal and ENDF/B-VIII.1 absorption (line) and with an earlier measurement from the Experimental Nuclear Reaction Data (EXFOR) database (crosses). Vanadium was used to validate the thermal neutron transmission capability that this work applies to the moderator materials.

Deliverables

• Submission of measured cross sections to the Experimental Nuclear Reaction Data (EXFOR) database: Planned.

• Submission of temperature-dependent TSL evaluations to the National Nuclear Data Center (NNDC): Planned.

• Publication describing IRMA, the open-source thermal scattering evaluation code developed as part of this work: Manuscript complete, in the ORNL review and release process.

• Publication describing the thermal neutron transmission capability developed at the VENUS beamline as part of this work: Manuscript complete, in the ORNL review and release process.

• Publication presenting the temperature-dependent measurements and TSL evaluations for SiC, ZrC, MgO, BeO, and Be₂C: In preparation.

Simplified CAD model of the SNTP critical assembly tank, grid plates, and fuel stalks.

In the late 1980s and early 1990s a series of critical experiments were performed at Sandia National Laboratories (Sandia) at the Sandia Pulse Reactor Facility (SPRF) in support of the Space Nuclear Thermal Propulsion Program (SNTP) with a total of 142 approaches to critical. The critical experiments used a particle bed type fuel containing TRISO—like 93% enriched UC fuel particles, graphite moderator particles, and zircaloy-4 filler particles. The particles were used to fill 19 fuel elements that were placed in a water filled tank that provided both moderation and reflection. These historic experiments will be documented and submitted to the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Technical Review Group (TRG) for entry into the ICSBEP Handbook of evaluated critical experiments.

Milestones and Deliverables

FY26: Documentation Retrieval

• Collect and organize historic documentation pertaining to SNTP experiments and document as Section 1 of the ICSBEP evaluation.

FY27: ICSBEP Evaluation

• Complete initial draft evaluation and internal and external reviews by end of January 2027.
• Submit draft of evaluation to FY27 ICSBEP TRG meeting.
• Attend FY27 ICSBEP TRG meeting and defend the evaluation.

Submit final evaluation for publication in August FY27. 

The IPEN/MB-01 New Plate-Type Core, from IEU-COMP-THERM-007 benchmark.

The Instituto de Pesquisas Energéticas e Nucleares (IPEN)MB01 reactor has recently been reconfigured with a new plate-type fuel core. The fuel elements are made of U₃Si₂-Al with an average density of 2.8 gU/cm³ and 19.75 wt% enriched in ²³⁵U. As pointed out in the EAW#1 topic area I.C.2, there is a lack of critical experiments at this enrichment level. Several critical experiments were already performed with this new core between 2020 and 2024, including different critical configurations with different fuel elements arrangements, with the control rods fully withdrawn or at specific positions, or adjustments of control rod position to find critical configurations at different temperatures between 20° and 80° C. The goal of the work performed is to create an ICSBEP benchmark from the performed experiments, and then to create a second evaluation with additional critical configurations, including the use of lower enriched fuel rods (to lower the average enrichment), and the insertion of graphite elements near the fuel plates.

Top view of designed critical experiments cases 1 (left, no graphite elements) and 6 (middle, 5 graphite elements), and concept of aluminum sleeved graphite element (right).

Milestones and Deliverables

FY24: Evaluation #1 work

• ORNL performs the internal and independent reviews of the new plate-type core for Evaluation #1 (E#1) by Monte-Carlo calculations.
  • Done.

FY25: ICSBEP evaluation #1 and design/planning of evaluation #2

• Submission of E#1 to the 2025 ICSBEP Technical Review Group (TRG) meeting.
  • Done.
• Defend E#1 at the ICSBEP TRG meeting in April 2025: 
  • Done. The evaluation was not accepted for publication and was re-submitted again at the 2026 ICSBEP TRG meeting.
• Design of additional critical experiments for Evaluation #2 (E#2): 
  • Done
• Critical experiments for E#2 are performed, and experimental data is obtained:
  • The experiments could not be performed, and E#2 is canceled.

FY26: ICSBEP evaluation #1 re-submission and ICSBEP guidance report

• ORNL writes a ICSBEP evaluation writing guidance and lessons learned report to help all future ICSBEP evaluators.
  • Done

The CROCUS reactor assembly with the PETALE reflector box shown in blue at the bottom of the picture.

In 2020, the Ecole Polytechnique Fédérale de Lausanne (EPFL, Switzerland) and the Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA, France) jointly conducted PETALE, a series of criticality experiments in the CROCUS reactor to validate the nuclear data of stainless steel and its main constituents (Fe, Ni, Cr). CROCUS is a well characterized zero power (100 W) teaching reactor with an existing reactor physics benchmark. Four configurations were measured for PETALE, using large reflectors of 304L stainless steel and its components (iron, nickel, and chromium). Criticality measurements as well as neutron transmission via activation foils at various depths in the reflector blocks. This project will benchmark these already completed experiments for inclusion in ICSBEP and to provide additional validation cases for isotopes related to structural materials. The project is a collaboration with between EPFL, the University of California, Berkeley, Lawrence Livermore National Laboratory, and Oak Ridge National Laboratory.

The MCNP geometry of the CROCUS reactor detailed model, showing the PETALE reflector box to the lower left of the fuel.

Milestones

• FY2026 Milestone: Complete benchmark draft and provide it to internal and external reviewers (Task 1)
• FY2027 Milestone: Complete internal and external review of benchmark (Task 2), and submit to the ICSBEP, defend at the ICSBEP Technical Review Group, and finalize the benchmark by making any required edits (Task 3)

Detailed SCALE model of polyethylene-reflected 3 by 3 UF6 cylinder array (left) with historical photograph of the actual experiment set-up prior to remote assembly (right).

Lawrence Livermore National Laboratory (LLNL), in partnership with Oak Ridge National Laboratory (ORNL), proposes to benchmark a series of experiments that were completed in 1973 at the Oak Ridge Critical Experiments Facility (ORCEF) but were never benchmarked as part of the ICSBEP. The experiments performed were array studies with bare, polyethylene-reflected and concrete-reflected configurations that were designed for nuclear criticality safety validation of transportation and storage configurations. The experiments used UF₆ cylinder model 8A containers and were filled with highly enriched uranium at 97.7% enrichment UF₆. The proposed benchmarks would fill an important gap as the ICSBEP does not currently have any UF₆ cylinder benchmarks.

Milestones and Deliverables:

FY25 and FY26: Complete benchmark draft and provide it to internal and external reviewers (Task 1)

FY27:  Complete internal and external review of benchmark (Task 2), and submit to the ICSBEP, defend at the ICSBEP Technical Review Group, and finalize the benchmark by making any required edits (Task 3)

Historical assembly drawing for Super Kukla, showing the stacked uranium places that comprised the machine.

Lawrence Livermore National Laboratory (LLNL) is benchmarking Super Kukla, a fast burst reactor operated by LLNL on the Nevada Test Site (now the Nevada National Security Site) from 1964 to 1978.  Super Kukla was designed to produce very large neutron pulses (10¹⁵ neutrons/cm²) on large samples (up to 36 inches high).  To allow both for a large inner sample cavity and ensure optimal thermal and shock performance, Super Kukla was constructed of 20% enriched uranium alloyed with 10% molybdenum.  The benchmark will fill an important gap as the ICSBEP does not currently have any high-assay low enriched uranium (HALEU) bare metal benchmarks, and this benchmark would likely become one of the fundamental benchmarks for nuclear data testing, similar to Lady Godiva (HMF-001) and Jezebel (PMF-001).

Milestones

•    FY2027 Milestone: Complete benchmark draft and provide it to internal and external reviewers (Task 1)
•    FY2028 Milestone: Complete internal and external review of benchmark (Task 2), and submit to the ICSBEP, defend at the ICSBEP Technical Review Group, and finalize the benchmark by making any required edits (Task 3)

Disassembly of the AGN-201M reactor at ISU in Summer 2025 (left) compared to the picture of the AGN-201M concept model with the creators of the reactor (right).

The AGN-201 reactor is a graphite reflected core comprised of a fuel system with UO­₂ powder enriched to approximately 20 wt% U-235 that is matrixed in polyethylene. The uranium polyethylene mixture is near optimal moderation minimizing the mass needed to be critical. Evaluating this reactor is useful for optimally moderated or graphite-reflected UO2 systems.  This project will disassemble the reactor to collect detailed information of the components, perform approach to critical experiments, and develop  an evaluation for submission to the ICSBEP.

Milestones and Deliverables

• Disassemble Reactor and collect measurements and samples on components for characterization
  • Completed June 2025
• Perform critical experiments of the reactor for evaluation inclusion to the ICSBEP
  • Ongoing
• Prepare and Submit Evaluation and Defend Evaluation at the TRG
  • April 2028

This project aims to design and execute criticality experiments that will serve as validation benchmarks to UF₆ criticality applications, conducted by teams from Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL). Current experimental facilities cannot support criticality experiments utilizing the UF₆ in a form and quantity needed to approximate the UF₆ transport containers. Thus, the criticality experiments in the ongoing project will act as surrogates, providing integral nuclear data validation coverage with high similarity (i.e., c_k) towards UF₆ 30B transport containers. Combination of high enriched uranium (HEU) plates, natural uranium (NU) plates, polyethylene sheets and reflectors, and C₂F₄ will be used to setup the experimental configurations at the National Criticality Experiments Research Center (NCERC). The aim is approximate the effective enrichment and the neutron spectrum of the UF₆ criticality applications, eventually providing validation coverage of UF₆ transportation with moderation exclusion and other fuel cycle criticality applications involving low moderated UF₆. The phases of the project span designing the experiments, material procurement, conducting the experiments, and benchmark evaluation for submission to the ICSBEP criticality handbook.

Milestones and Deliverables

• Finalize Experiment Designs: October 2026
• Procure Components needed for the experiment: December 2026
• Perform Experiments: May 2027
• Present Evaluation to the ICSBEP TRG: March 2028