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DOE OSTI2016

Needs and Requirements for Future Research Reactors (ORNL Perspectives)

The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Nuclear Science User Facilities Irradiation Capabilities at Oak Ridge National Laboratory

The Nuclear Science User Facilities (NSUF) within the Department of Energy (DOE) Office of Nuclear Energy offer researchers the opportunity to access world-class nuclear facilities at no cost to the user. Oak Ridge National Laboratory (ORNL) offers users access to the High Flux Isotope Reactor (HFIR) in addition to a wide variety of post-irradiation examination facilities. The ORNL irradiation engineering team has compiled a large number of successful designs for irradiating nuclear fuel and material specimens. NSUF users can leverage these existing designs to significantly reduce the time, cost, and risk of failure when designing new irradiation experiments. This work summarizes some of the irradiation experiment designs and capabilities that have been developed at ORNL in the hope that this information will help improve future NSUF proposals and inspire new ideas for irradiations that can be performed in the HFIR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Conceptual Design Parameters for HFIR LEU U-Mo Fuel Conversion Experimental Irradiations

The High Flux Isotope Reactor (HFIR) is a versatile research reactor that is operated at the Oak Ridge National Laboratory (ORNL). The HFIR core is loaded with high-enriched uranium (HEU) and operates at a power level of 85 MW. The primary scientific missions of the HFIR include cold and thermal neutron scattering, materials irradiation, and isotope production. An engineering design study of the conversion of the HFIR from HEU to low-enriched uranium (LEU) fuel is ongoing at the Oak Ridge National Laboratory. The LEU fuel considered is based on a uranium-molybdenum alloy that is 10 percent by weight molybdenum (U-10Mo) with a 235 U enrichment of 19.75 wt %. The LEU core design discussed in this report is based on the design documented in ORNL/TM-2010/318. Much of the data reported in Sections 1 and 2 of this document was derived from or taken directly out of ORNL/TM-2010/318. The purpose of this report is to document the design parameters for and the anticipated normal operating conditions of the conceptual HFIR LEU fuel to aid in developing requirements for HFIR irradiation experiments.

Chandler, David↗

DOE OSTI2026

Investigation of irradiation damage and heat deposition: a comparative analysis for HEU-to-LEU conversion in HFIR

The planned conversion of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel requires detailed evaluation of experiment-relevant parameters to ensure continued performance for materials testing and isotope production. Here, this study presents the first comprehensive assessment of displacements per atom (dpa) and heat deposition rates in target materials within the HFIR flux trap with both HEU and candidate LEU core configurations. Seven analyses were conducted to evaluate key performance metrics, including fast neutron flux distribution, cross section response functions, cross section data, and local dpa and heat deposition rates using mesh- and cell-based tallies. Simulations employed Shift, Monte Carlo N-Particle (MCNP), and the HIFR Controller (HFIRCON) tool suite for high-fidelity transport and depletion modeling. The LEU designs—using U 3 Si 2 -Al dispersion fuel and operating at 95 MW—were compared to the current 85 MW HEU configuration. Results show that while the candidate LEU cores exhibit higher dpa rates due to a harder spectrum and extended cycle lengths, they also demonstrate reduced heat deposition rates in irradiation experiments, primarily due to increased gamma self-shielding from higher 238 U content in the core. These findings confirm that LEU conversion can maintain HFIR’s materials irradiation capabilities but may require redesigning existing experimental hardware.

HEU↗

DOE OSTI2017

Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannah River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H 2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H 2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H 2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Criticality Scoping Calculations for a Multi-Function Dissolver Insert

As one of its missions, the Savannah River Site seeks to be a leader in the safe and productive disposition of special nuclear materials. One of the premier programs is the disposition of used nuclear fuel, primarily from research reactors, through dissolution, purification, and down blend into low enriched uranium for commercial power reactors. Prior to use, this fuel contained highly enriched uranium (HEU). In an effort to advance this expertise, this study has been performed to demonstrate the feasibility to disposition a number of special geometry HEU fuels. These fuels come from a number of different domestic and foreign research reactors which had geometries considerably different from round or square bundles of plates or rods. As such, they cannot be dispositioned in current H-Canyon dissolver inserts so a new, multi-function dissolver insert may be considered. While this is not a current mission, it is believed that this study's scoping calculations demonstrate the feasibility of the site taking such a mission. At the H-Canyon facility, a dissolver is a large stainless steel pot, loaded with a nitric acid solution into which uranium-alloy fuel is dissolved. The dissolver uses a device called an insert to feed the fuels to the acid solution in a controlled manner. Most fuel forms fit into one of the existing inserts which include one with ten {approx}5-inch diameter circular wells, one with four {approx}4.5 inch by {approx}15 inch slab wells, and one specifically designed for the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). This study considered five bounding fuel candidates for scoping the potential new insert. The fuels include a generic annular plate fuel (APF), the Tower Shield Reactor (TSR) fuel elements, the Reactor A Haute Flux (RHF) core, Knolls Atomic Power Laboratory (KAPL) reactor fuel, and SLOWPOKE cores. These are expected to bound a range of proposed fuel candidates. Scoping studies were made to establish geometric restrictions for the new dissolver insert and to determine if these fuels could safely be dissolved. Homogenous models were used for the scoping study due to time constraints. Computational modeling was performed using KENOVI in the SCALE 6.1 code package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Assessing the Impact of Large Removable Beryllium Reflector Experiments on HFIR Performance and Safety Metrics

This study investigated the effect of various removable beryllium (RB) experiment configurations on High Flux Isotope Reactor (HFIR) metrics to address increasing interest in these facilities for materials and fuels irradiation research. The RB reflector contains eight large and four small irradiation experiment facilities that offer excellent neutron flux conditions to perform fission and fusion reactor materials and fuels irradiation research. This work aims to outline acceptable RB configurations based on safety, performance, and programmatic metrics. This study assessed experiment effects on reactivity, cycle length, fission rate density distributions, and neutron flux distributions using the Shift, HFIRCON, and SCALE ORIGEN codes. Initial evaluations focused on generic experiment materials including aluminum plugs, stainless steel plugs, molybdenum plugs, and aluminum plugs with gadolinium shields. Subsequent analyses of MiniFuel experiments were performed to evaluate a heterogenous experiment, consisting of a more complex geometry and bearing several materials, and to test the correlations developed with the generic materials on a real experiment. Furthermore, the effects of neutron poison concentrations in the standard beryllium plugs were evaluated. When assuming a reference RB configuration with eight large fresh beryllium plugs, perturbed configurations with three aluminum plugs, one stainless steel plug, one molybdenum plug, one gadolinium-shielded aluminum plug, and one MiniFuel experiment resulted in a cycle length reduction of less than 1.3 days, the current threshold before requiring additional approvals. Configurations with five aluminum plugs, one stainless steel plug, one molybdenum plug, one gadolinium-shielded aluminum plug, and two MiniFuel experiments meet the 1.3 day limit if irradiated beryllium plugs are considered. Fuel element fission rate density distributions remained within safety limits, with maximum local increases under 9%. Neutron flux calculations revealed large thermal flux depressions inside and around the perturbed RB facilities, while epithermal and fast neutron fluxes increased because of reduced neutron moderation by the perturbed materials. These findings provide valuable guidance for optimizing RB configurations to balance safety and performance at HFIR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

The suite of small-angle neutron scattering instruments at Oak Ridge National Laboratory

Oak Ridge National Laboratory is home to the High Flux Isotope Reactor (HFIR), a high-flux research reactor, and the Spallation Neutron Source (SNS), the world's most intense source of pulsed neutron beams. The unique co-localization of these two sources provided an opportunity to develop a suite of complementary small-angle neutron scattering instruments for studies of large-scale structures: the GP-SANS and Bio-SANS instruments at the HFIR and the EQ-SANS and TOF-USANS instruments at the SNS. This article provides an overview of the capabilities of the suite of instruments, with specific emphasis on how they complement each other. As a result, a description of the plans for future developments including greater integration of the suite into a single point of entry for neutron scattering studies of large-scale structures is also provided.

47 OTHER INSTRUMENTATION↗

DOE OSTI2012

Calculation of heating values for the high flux isotope reactor

Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Calculation of Heating Values for the High Flux Isotope Reactor

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

ATF Cladding Mechanical Properties Report: Capability Demonstration

This report documents mechanical testing capability demonstration activities performed in fiscal year (FY)2025 at Oak Ridge National Laboratory(ORNL)on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~13GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide(SiC)thermometry (TM) was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target. Mechanical testing at ambient temperature demonstrated the expected irradiation-induced hardening, with yield strength(YS)and ultimate tensile strength(UTS)values increasing substantially relative to unirradiated counterparts. However, this strengthening was accompanied by a reduction in ductility, as indicated by lower uniform and total elongations(UE and TE). Both coated and uncoated claddings exhibited similar mechanical response, though Cr-coated specimens showed surface cracking perpendicular to the loading direction, attributable to the hardness mismatch between the coating and substrate. Fracture in all cases remained ductile, and no coating spallation was observed following HFIR irradiation. Complementary efforts were directed toward the fabrication of test specimens from commercially irradiated cladding (rod 47I, ~31.1 GWd/t average burnup). Axial sectioning and computer numerical control(CNC)machining successfully produced ATT geometries suitable for benchmarking against HFIR-irradiated specimens. This capability enables direct comparison of cladding behavior between test reactor and commercial reactor environments, thereby supporting the validation of HFIR as a surrogate irradiation platform for accident tolerant fuel (ATF) development. Once HFIR irradiations are completed in FY26,the relevant comparison tests will be completed. Collectively, the FY 2025 PIE campaign has provided mechanical performance data for irradiated advanced claddings The demonstrated capabilities support the framework for mechanical testing and further evaluations in subsequent years. These efforts will represent an important contribution toward the licensing and deployment of Cr-coated zirconium alloy cladding as a near-term ATF solution.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Building Interfaces with Echo Using R

The following report details the functionality that was developed to directly query the MongoDB and read in echo records from within R, and the ability to create a cinema database directly from echo records. Both of these efforts used the data collected under the MINOS project for development. The nuclear facility under consideration is the High-Flux Isotope Reactor (HFIR) and co-located Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory in Oak Ridge, TN. HFIR is an 85 MW research reactor and is used primarily for production of a medical radioisotopes, material irradiation experiments, neutron activation, and neutron scattering. Targets for the reactor are fabricated and processed and dissolved at REDC, in addition to other glove-box and hot-cell type activities. An overhead view of the facility is provided. The rest of the report is organized as follows: Section 2 describes how to query the MongoDB and load in echo records from within R, with background information on the database type, MongoDB, and the file format of echo records, hdf5, as well as lessons learned during the development phase. Section 3 details the functionality and usage of creating a cinema database directly from echo records within R. A series of appendixes display all the R code (Section 5), some example scripts where the code is used (Section 6), and useful links for further exploration (Section 7).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

MINOS Infrasound Analysis Synopsis

This report was written as a guide to working with infrasound data collected as part of the Multi-Informatics for Nuclear Operations Scenarios (MINOS)project, an NA-22 funded venture. The main purpose of overall MINOS project is the combination of multiple, disparate data modalities to characterize the operations at a nuclear facility, specifically instrumenting and studying the High-Flux Isotope Reactor (HFIR) and Radiochemical Engineering Development Center (REDC) locate at Oak Ridge National Laboratory in Oak Ridge, TN. HFIR is an 85 MW research reactor and is used primarily for production of medical radioisotopes, material irradiation experiments, neutron activation, and neutron scattering. Targets for the reactor are constructed, processed, and dissolved at REDC. REDC also hosts other glove-box and hot-cell type activities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

High Flux Isotope Reactor Core Analysis-Challenges and Recent Enhancements in Modeling and Simulation

A concerted effort over the past few years has focused on enhancing the core depletion models for the High Flux Isotope Reactor (HFIR) as part of a comprehensive study for designing a HFIR core that would use low-enriched uranium (LEU) fuel. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed for use as a reference for the design of an LEU fuel for HFIR and to improve the basis for analyses that support HFIR’s current operation with high-enriched uranium (HEU) fuel. This paper summarizes the recent improvements in modeling and simulation for HFIR core analyses, with a focus on core depletion models.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2012

Performance and Safety Parameters for the High Flux Isotope Reactor

A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Sensitivity Coefficients for Diffusion Coefficients and Other Reactor Physics Parameters using CE TSUNAMI-3D

The TSUNAMI (Tools for Sensitivity and Uncertainty Analysis Methodology Implementation) capabilities within the SCALE code system leverage sensitivity coefficients for many criticality safety applications, including quantifying the data-induced uncertainty in the eigenvalue of critical systems, quantifying computational biases, and selecting applicable benchmark experiments. TSUNAMI sensitivity and uncertainty analysis tools have traditionally been limited to criticality safety application spaces, but the introduction of Generalized Perturbation Theory (GPT) sensitivity methods by Jessee in 2009 and Perfetti in 2013 has extended the scope of SCALE adjoint-weighted sensitivity and uncertainty analysis to include reactor physics applications. The continuous-energy (CE) TSUNAMI-3D GPT sensitivity analysis methods were originally developed to calculate sensitivity coefficients for a limited number of reaction-rate responses using CE Monte Carlo methods. These tools were recently applied in an Oak Ridge National Laboratory (ORNL)-directed research and development project to improve the efficiency of {sup 252}Cf production in the High Flux Isotope Reactor (HFIR) by understanding the sensitivity of key heavy actinide capture-to- fission ratios to potential system design changes. This work further extends the applicability of the CE TSUNAMI- 3D GPT methods to reactor physics applications by enabling the calculation of sensitivity coefficients for an expanded set of reaction rate responses-including a first-ever capability for conducting sensitivity analysis of neutron diffusion coefficients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Development of a Scale Model for High Flux Isotope Reactor Cycle 400

The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Comparison of HFIR Core Hydraulic Models with Operational Data

The core flow channels of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) have been modeled with three different computer codes and calculated results have been compared with available plant operational data. The three codes employed are RELAP5-3D, TRACE, and COMSOL. This work was performed to develop consistent boundary conditions for three-dimensional (3-D) computational fluid dynamics (CFD) models of the HFIR core. The variables of interest for the CFD models include: (1) the inlet velocity and (2) outlet pressure for each coolant channel, and (3) the axial pressure difference between narrow and wide adjacent channels. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗