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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

AFC FY 2023 HFIR Irradiation Test Matrix – Supported by the Design of a Ring Specimen Irradiation Vehicle

The development and deployment of accident-tolerant cladding materials in light water nuclear reactors requires the collection of neutron-irradiated properties of these materials with specific consideration to the material properties orientation (i.e., tube geometry). To that end, a test plan to irradiate several cladding tube and ring specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was proposed, along with the development of a new capsule design that features ring specimens with a preexisting gauge section. This ring specimen geometry removes the need to machine a gauge section on the irradiated specimen and thus allows for direct application of a ring-pull test post-irradiation. The test matrix includes uncoated and chromium-coated Zircaloy-4 and advanced zirconium alloy cladding specimens, pre-characterized at ORNL. Plans are to irradiate the specimens at about 4 and 10 dpa—corresponding to two and five cycles at the HFIR centerline—with a target specimen gauge section temperature of 330°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

HFIR Activity Workbook Generator (HAWK) User Guide

The HFIR Activity WorkbooK generator (HAWK) is a Python code that automates and streamlines the activity calculation of samples after irradiation in the High Flux Isotope Reactor (HFIR). HAWK’s results provide estimates of the activity and nuclide inventory of irradiated specimens before they are moved to hot cell facilities, where they undergo post-irradiation examination. The samples’ activity results guide the packing of shipping containers and inform the accountable inventories for the hot cell facilities. The toolkit was originally developed by Charles Daily, a former R&D staff member at Oak Ridge National Laboratory (ORNL). As of May 2025, HAWK is developed by the Radiation Transport & HPC Methods Group (Nuclear Energy and Fuel Cycle Division) at ORNL. Figure 1 presents HAWK’s workflow. To use HAWK, users need to: 1. Develop an Excel input workbook (i.e., XLSX extension) containing data from the experiment’s materials, irradiation history (cycles), and irradiation positions. 2. Make minor edits to an existing template JSON file (i.e., auxiliary_data.JSON) and to the Python driver. The driver sets the necessary environment variables, defines the material compositions, and ultimately calls HAWK. Once configured, HAWK runs the Oak Ridge Isotope Generation code (ORIGEN) to calculate the masses, activities, and heat load at the end of irradiation for each isotope in the specimen. ORIGEN is part of SCALE, ORNL’s in-house computational tool for performing nuclear safety and design calculations. Following this step, HAWK postprocesses the results and generates three output workbooks summarizing the activity calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Ray-Tracing Simulations Characterizing the Performance of the Proposed 2028 HFIR HB4 Main Shutter

The Main Shutter at HB4 will serve two purposes after the HFIR Beryllium Reflector Replacement, planned to take place in 2028. First as the primary certified safety control permitting the passage of neutrons from the cold source in the HFIR pressure vessel into the cold guide hall, and second as the first set of reflecting surfaces used to guide neutrons from the source and into the individual guide starts for each instrument in the cold guide hall. This report outlines a the simulations and analyses used to conclude that the HB4 Main Shutter Insert alignment is critical to associated instrument performance. In general, the alignment of these surfaces to within +/-0.25 millimeters and +/-0.05° across all dimensions will ensure losses are no worse than 1% for any of the guide starts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Long-Length SiC/SiC Tube Irradiation Experiment at LWR Temperatures Inserted into HFIR

This report describes the pre-irradiation characterization and successful assembly of a High Flux Isotope Reactor (HFIR) irradiation experiment designed to assess radiation-induced lateral bowing of silicon carbide fiber–reinforced, silicon carbide matrix composite (SiC/SiC) components under a radial fast neutron flux gradient at representative light-water reactor (LWR) temperatures of approximately 300 °C. The over-arching goal of this work is to validate thermomechanical models of SiC/SiC composite fuel assembly distortion used to predict the extent of lateral bowing and dose at which bowing reaches a maximum value. Ultimately, these models are used to ensure that fuel assembly distortion does not block coolant channels, interfere with control rod/blade movements, and/or impact local reactivity. The experiment contains six cladding tube specimens that are approximately 558.5 mm long with representative pressurized water reactor (PWR) outer and inner diameters of 9.5 mm and 7.5 mm, respectively. The specimens were thoroughly characterized prior to irradiation using traditional dimensional inspection and surface profilometry so that these measurements can later be compared with similar measurements that will be made post-irradiation to determine radiation-induced deformations. Furthermore, fine engraving markers were inscribed along the outer surfaces of the specimen and mapped using analytic photography and a three-dimensional stage. This technique allowed for accurate measurements of the marker spacings, which can be compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. A light curtain optical micrometer was also employed to measure the pre-irradiation bowing of each specimen. The experiment was successfully assembled and inserted into position VXF-19 for HFIR cycle 516, which began May 12, 2026. The experiment will be removed after cycle 519, which is currently scheduled to end in December 2026.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

HFIR Steady State Heat Transfer Code (HSSHTC) Statistical Uncertainty Analysis

HSSHTC, the safety basis steady state TH code for HFIR, uses a highly conservative approach in which all input and calculation uncertainties are resolved simultaneously at their most limiting setting. This results in excessive conservatism which does not account for the high unlikelihood of such simultaneous worst-case conditions. The present study explores an alternative approach, BEPU, in which reasonable working assumptions for the probability distribution of each input uncertainty are used to determine a relationship between burnout power margin and core fuel failure probability. This was performed under a philosophy of perturbing uncertainty parameters already defined within the HSSHTC methodology while preserving the HSSHTC calculation approach and solution methodology itself. Based on the assumptions employed in this study, the BEPU approach resulted in a 0.29 increase in burnout power ratio (25 MW increase in burnout power) compared to the latest HSSHTC calculations of C-HFIR-2026-004. The study can be refined in the future by employing fuel fabrication data to provide more realistic input distributions. Future changes to the HSSHTC methodology would potentially allow a more comprehensive treatment of uncertainties which may further increase the burnout power ratio.

Wysocki, Aaron [ORNL] (ORCID:0000000222043779)↗
DOE OSTI2020

Materials Data on HfIr by Materials Project

HfIr is Tetraauricupride structured and crystallizes in the cubic Pm-3m space group. The structure is three-dimensional. Hf is bonded in a body-centered cubic geometry to eight equivalent Ir atoms. All Hf–Ir bond lengths are 2.84 Å. Ir is bonded in a body-centered cubic geometry to eight equivalent Hf atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on HfIr by Materials Project

HfIr is alpha-derived structured and crystallizes in the monoclinic P2_1/m space group. The structure is three-dimensional. Hf is bonded in a 7-coordinate geometry to seven equivalent Ir atoms. There are a spread of Hf–Ir bond distances ranging from 2.69–2.88 Å. Ir is bonded in a 7-coordinate geometry to seven equivalent Hf atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on HfIr by Materials Project

HfIr is beta-prime cadmium gold structured and crystallizes in the orthorhombic Pmma space group. The structure is three-dimensional. Hf is bonded in a distorted body-centered cubic geometry to eight equivalent Ir atoms. There are six shorter (2.81 Å) and two longer (3.02 Å) Hf–Ir bond lengths. Ir is bonded in a distorted body-centered cubic geometry to eight equivalent Hf atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO2), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Effects of HFIR neutron irradiation on fracture toughness properties of standard and Ni-doped F82H

F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. Here, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58 Ni, and F82H doped with 1.4% 60 Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Black body grids at HFIR CG-1D and NCNR BT2

Data sets from evaluation of prototype black body grids produced by thick film printing at the HFIR CG-1D cold neutron imaging beamline and the NCNR BT2 thermal neutron imaging beamline.

black body grids↗
DOE OSTI2020

A Replacement Cold Neutron Guide System for HFIR (Conceptual Design Report)

The design of a replacement cold neutron guide system for the High Flux Isotope Reactor (HFIR) is presented. The proposed six-channel design uses the latest technologies and is optimally designed to account for operational and site constraints. These include the design of the cold source and the beam tube, the size and shape of the penetrations in the reactor confinement boundary, and the shape and orientation of the guide hall. The existing user instruments are a macromolecular diffractometer (IMAGINE), an IMAGING station, two small-angle neutron scattering (SANS) instruments (Bio-SANS and GP-SANS), and a cold triple-axis spectrometer (C-TAX). These instruments will all be reinstalled at end-of-guide positions with improved performance, and a position for a new high-performance neutron spin echo instrument will be created. The proposed guide design requires an extension to the guide hall toward the south. The calculated flux gain for a typical experiment is between a factor of 1.3 and 3 for all instruments, with the exception of the C-TAX instrument, which stands to gain a factor of >30 in flux at the sample. The new guide system will also offer additional performance gains, such as an increase in Q-range on Bio-SANS. The stated performance gains do not take into account any potential performance gains resulting from renewal and realignment of the guide system. The design of the guide system in the common sections up to about 25 m from the source is considered to be complete and ready for engineering design, while the designs of the instrument-specific guides are in some cases still a work in progress. The present report describes the current status of those designs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Final Status Report Confirming Completion of Mechanical Testing of HFIR Irradiated Nanocomposite Materials

This report provides a final status report on the in-cell mechanical testing of High Flux Isotope Reactor (HFIR)-irradiated nanodispersion-strengthened materials at the Irradiated Materials Examination and Testing (IMET) facility. All tension tests were performed at room temperature with a nominal strain rate of 0.018 mm/mm/min using shoulder-loading grip sets by following the standard testing procedure in ASTM E8/E8M. Round 1 (unirradiated) and round 2 (0.7 dpa) testing was completed in FY 2021. This report confirms completion of round 3 (1.4 dpa) and round 4 (2.1 dpa) tensile testing in FY 2022 with examples of test data collected on select sample conditions. Additionally, the report includes IDs of samples selected for shipment to Idaho National Laboratory (INL) for additional characterization.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Assessment and Planning of HFIR Test Articles

Burnable absorbers (BAs) are introduced into nuclear fuels to aid in controlling the activity of the irradiated fuel at the beginning of life in a light water reactor (LWR). This improves the efficiency of the reactor since it allows for higher enrichments (> 5 % 235 U) to be used in a LWR, which increases the burnup limit and therefore the cycle length. The purpose of the BA is to absorb neutrons to prevent power peaking, and the concentration of BA is consumed under irradiation. In order to understand the effect of the BAs on the microstructure and thermophysical properties of the fresh and irradiated fuel, neutron irradiation tests will be carried out at the High Flux Isotope Reactor (HFIR). These tests will include Gd 2 O 3 doped UO 2 fuels with concentrations that vary from 4 to 10 weight % Gd 2 O 3 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

As-Run Neutronic Calculations of HFIR Mini-Fuel Experiments: Guidelines for Analysts

This report provides guidelines for analysts performing radiation transport calculations for mini-fuel experiments irradiated in the High Flux Isotope Reactor (HFIR). Procedures, practical recommendations, and modeling conventions which are essential to ensuring consistency and high fidelity in as-run simulations have been compiled. The content is intended to be a portion of the official training material for new staff in this area, offering both technical references and practical guidance derived from previous experience. This report is a living document and will be periodically updated to incorporate new developments, methodological improvements, and lessons learned from ongoing and future irradiation campaigns.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO 2 ), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗