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DOE OSTI2016

Dissolution flowsheet for high flux isotope reactor fuel

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 and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2 ) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Spallation Neutron Source Second Target Station Integrated Systems Update

The Spallation Neutron Source (SNS) was designed from the beginning to accommodate both an accelerator upgrade to increase the proton power and a second target station (STS). Four workshops were organized in 2013 and 2014 to identify key science areas and challenges where neutrons will play a vital role [1-4]. Participants concluded that the addition of STS to the existing ORNL neutron sources was needed to complement the strengths of High Flux Isotope Reactor (HFIR) and the SNS first target station (FTS). To address the capability gaps identified in the workshops, a study was undertaken to identify instrument concepts that could provide the required new science capabilities. The study outlined 22 instrument concepts and presented an initial science case for STS [5]. These instrument concepts formed the basis of a planning suite of instruments whose requirements determined an initial site layout and moderator selection. An STS Technical Design Report (TDR) documented the STS concept based on those choices [6]. Since issue of the TDR, the STS concept has significantly matured as described in this document.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

High y + Shear-Stress Turbulence Implementation for High Flux Isotope Reactor Narrow Channel Flows

The research objective of this work was to improve the engineering predictions of the turbulence characteristics of flows in curved narrow channels. Such channel flows are commonly encountered in nuclear research and test reactors, with one of them being the high-flux isotope reactor (HFIR). Research reactors bear high heat fluxes, and the proper computing of turbulence is paramount for safe and reliable reactor operation. The study builds on the results of a previous direct numerical simulation of turbulence to inform a well-known Reynolds-averaged Navier–Stokes shear-stress turbulence model and improves its accuracy in simulating parallel channel flows. A new formulation of the loss term in the dissipation conservation equation is suggested. Combined with high wall distance computational grids, the new implementation provides a fast-running flow solution, suitable for engineering purposes. Model generalization for parallel channel flows, in a broader range of frictional Reynolds numbers, is suggested by introducing a new form of the model constants.

CFD↗

DOE OSTI2016

Comparison of Calculated and Experimental Results for a Boiling/Condensing Experimental Facility

A new experimental facility is being developed for materials irradiation and testing at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). Details of this facility have been presented before. A prototype of this facility, the Thermosyphon Test Loop (TSTL) has been built, and experimental data have been obtained and analyzed. Pretest calculations for this facility with the RELAP5-3D code have been presented previously as well as other calculations with the TRACE code. The results of both codes were very different. RELAP5-3D predicted much higher pressures and temperatures than TRACE. This paper compares calculated results with the TSTL experimental data. Comparison of calculations with the codes RELAP5-3D and TRACE with experimental data of the new TSTL facility has shown that TRACE results agree well with the data and that RELAP5-3D calculates very high pressures and temperatures. The TRACE code is well suited to model this facility and is being used for future calculations. (authors)

42 ENGINEERING↗

DOE OSTI2016

Experimental Data and Analysis for a New Irradiation Facility

A new experimental facility is being developed for materials irradiation and testing at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). Details of this facility have been presented before. A prototype of this facility, the Thermosyphon Test Loop (TSTL), has been built, and experimental data have been obtained and analyzed. The purpose of the tests was to establish a safe operating envelope for actual fueled experiments in a HFIR experimental facility. Both steady-state and transient experiments were conducted. The data will also be used to validate computer simulations of the thermosyphon so that those codes can be used for future safety-basis calculations. This paper presents a summary of the TSTL experimental data and analysis. A report with all the data and analyses will be published in the near future. Experimental data from 51 tests in the TSTL have been obtained. This extensive set of high quality data can be used to benchmark codes with boiling and condensing phenomena. This new proposed irradiation facility allows materials to be irradiated without concern for HFIR coolant contamination (from specimen failures) as it uses a separate coolant, that is isolated (except thermally) from the HFIR primary coolant. (authors)

42 ENGINEERING↗

DOE OSTI2016

Neutron Irradiation Effects on the Graphitic Structure of a Fine-Grain Graphite

Multiple programs are being performed worldwide that investigate the neutron irradiation-induced property changes in graphite. Similar programs have been performed for over 60 years, and still there is little understanding about why different graphite grades respond differently to the same irradiation conditions. The current thought is that there are multiple microstructural features that evolve during neutron irradiation and different features control the different physical property evolutions. Multiple neutron irradiation campaigns are underway at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR). All of these campaigns utilize a comprehensive irradiation matrix that will allow investigation of the physical and microstructure changes, and will provide insight into which microstructure changes are controlling different physical property changes. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2015

Safe solution dissolution of high flux isotope reactor cores - 14490

The H-Canyon facility at the Savannah River Site is the U.S.A.'s only remaining full-scale chemical separations plant. H-Canyon maintains the capacity to safely and efficiently disposition a large inventory of excess nuclear material from across the Department of Energy (DOE) complex and used nuclear fuel from foreign and domestic research reactors. One such fuel that H-Canyon may dissolve is cores from the High Flux Isotope Reactor (HFIR). By employing realistic analysis, the need for administrative controls was eliminated. A new criticality modeling approach has eliminated the need for administrative controls for HFIR dissolution operations in H-Canyon. The primary upsets of concern are an over-mass event and an overconcentration event. Several initiating events were considered for each type of upset. Over-concentration events were deemed to be not credible by taking credit for the chemistry of acid consumption, i.e. there are not enough moles of acid to dissolve enough moles of uranium in order to reach a critical configuration. Over-mass events were deemed to be not credible by taking credit for the annular design of the fuel elements. By using realistic analysis, both types of upsets have been shown to be subcritical, and therefore administrative controls are not needed to prevent an inadvertent criticality. In this paper, the new methodology for performing criticality analyses for dissolving HFIR fuel elements will be presented, and the elimination of administrative controls will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Additive Manufacturing of Research Reactor Control Elements and Subsequent Neutron Irradiation

Additive manufacturing techniques have made major strides in recent years and are now capable of creating complex components in near-final form. The nuclear industry has many applications with complicated and even unique components that are time consuming, expensive, and more difficult to make using traditional fabrication routes. Modern fabrication techniques possess the potential to reduce fabrication lead-time and costs while increasing quality and efficiency. However, these methods have not been adopted due to the need for strict validation for nuclear applications. Unlike structural components that largely consist of a uniform material, for nuclear reactor core components incorporation of fissile or neutron poison materials inside a cladding becomes necessary. This in turn requires a more complex and flexible manufacturing method to produce these hybrid components. In this study, demonstration of ultrasonic additive manufacturing (UAM) process to produce components resembling research reactor control elements was carried out]. Specifically, the control elements for ORNL's High Flux Isotope Reactor (HFIR) were targeted. These control elements (consisting of an inner cylinder and four outer plates) are placed on the periphery of the circular HFIR core. The control elements consist of Al6061-T6 alloy with a 13-cm-long grey and a 56-cm-long black region, corresponding to 40 wt% of Ta and 30 wt% Eu{sub 2}O{sub 3} (europia) are incorporated along their length. In this study various fabrication parameters for production of aluminum blocks with and without embedded Eu{sub 2}O{sub 3} and Ta bearing compacts were explored. Select optimized specimens we then irradiated inside HFIR for one reactor cycle. The UAM process exploits ultrasonic energy to facilitate solid-state welding of metal foils in contact under pressure. This additive process is carried out in successive layers to build up the material and when coupled with CNC milling, it yields complex 3D components. The milling and subsequent additive steps may be used to embed foreign material into the material undergoing buildup. In this manner fissile or neutron poison elements may be incorporated into the complex structure, though in an inhomogeneous fashion. Multiple Al-6061 blocks were produced by UAM using 25 mm wide by 0.15 mm thick Al-6061-H18 tapes. The mechanical properties of these blocks in the as-fabricated condition was characterized in detail and it was shown that a great degree of anisotropy in the mechanical behavior exists between the direction in the tape plane and the direction perpendicular to the tapes (buildup direction) The latter exhibited reduced modulus and significantly smaller strength. Post production heat treatment appeared to greatly enhance the strength in the buildup direction. Multiple tensile specimens machined per SS-J3 geometry were prepared for irradiation testing. The tensile specimen matrix was set up to enable post irradiation examination of mechanical properties in various directions as well as after heat treatment. A small block with embedded Ta- and Eu{sub 2}O{sub 3}-bearing compacts, three of each kind, was also produced, as demonstrated previously. These compacts were 5.7 mm in diameter and 1.6 mm in height and contained 60 vol Ta or Eu{sub 2}O{sub 3}. The poison-bearing Al-6061 block alongside the tensile bars was loaded into a perforated rabbit (irradiation capsule) for irradiation in HFTK. The perforated rabbit allows penetration of HFIR water coolant into the Interior of the capsule in direct contact with the test specimens. The rabbit was loaded into the flux trap region at the uppermost position in the core. Irradiation temperature was 68 deg. C and fast neutron (E>0.1 MeV) fluence was 10{sup 25} neutrons/m{sup 2}. Currently PIE activities are ongoing that are focused on mechanical testing of irradiated tensile bars and metallographic examination of neutron poison bearing block.

36 MATERIALS SCIENCE↗

DOE OSTI2026

Design Status and Experimental Strategy for Initial Molten Salt Irradiation Experiments in HFIR

The development and deployment of molten salt reactor (MSR) technologies require experimental capabilities that can evaluate molten fuel salt behavior and structural material performance under representative irradiation conditions. Although modeling and separate effects testing provide important insight, there remains a critical lack of in-pile data that capture the coupled effects of neutron irradiation, temperature, salt chemistry, and time. Informed by lessons learned from historical MSR programs and recent international irradiation efforts, this report presents a structured approach to addressing existing gaps in molten salt irradiation testing using the High Flux Isotope Reactor (HFIR). A phased irradiation strategy is presented that incrementally increases experimental complexity while managing cost, risk, and facility constraints. The framework progresses from passive, nonfueled static capsule experiments to fueled, instrumented, and ultimately circulating molten salt systems, providing a pathway for materials screening, mechanistic understanding, and qualification-relevant testing. The report defines a near-term Phase 1 passive capsule concept, associated irradiation conditions, and a conceptual post-irradiation examination strategy. Ongoing modeling, out-of-pile testing, and safety analysis activities are described to support continued capability maturation and to enable future phases of molten salt irradiation testing in support of MSR technology deployment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Neutron fluxes in test reactors

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

HFIR Core Modeling and Analysis for the PROSPECT Reactor Antineutrino Experiment

Nuclear reactors have been employed as a copious source of antineutrinos for over six decades. Research initially directed at the study of antineutrino physics has led to applications of these highly penetrating particles to reactor safeguards because the antineutrino flux and spectrum can be related to the state of the reactor core, including burnup evolution. One of the main challenges of antineutrino-based safeguards is operation of the detectors near surface (within the reactor containment or in the close proximity). Lack of overburden results in increased contribution to the detector noise from cosmic radiation. Similarly to safeguards, experiments to definitively confirm or exclude the existence of sterile neutrino oscillations require detector technologies capable of operating at short baselines (tens of meters) at the surface. The Precision Reactor Oscillation and Spectrum Experiment (PROSPECT) at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is an experiment with a main objective to make a precision measurement of the flux and energy spectrum of antineutrinos emitted from nuclear reactors in order to search for the oscillation signature of sterile neutrinos. Fissions of highly-enriched uranium of HFIR will serve as a source of antineutrinos. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Depletion Analysis of Additively Manufactured Control Elements in HFIR

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory is a flux-trap type research reactor which provides one of the highest steady-state neutron sources in the world. HFIR neutrons are used for a range of research applications, including scattering experiments, materials irradiation, and isotope production. The HFIR core operates at 85 MW and is light water cooled and moderated. The core consists of several concentric cylindrical regions, as depicted in Fig. 1. At the center of the core is the flux trap, where the thermal flux reaches a maximum on the order of 10{sup 15} neutrons/cm{sup 2}-s. Immediately surrounding the flux trap are two annular fuel elements, an inner fuel element and an outer fuel element, each loaded with hundreds of thin fuel plates curved into an involute shape. Surrounding the outer fuel element are two thin control elements (CEs), an inner CE (ICE) and an outer CE (OCE). The CEs are axially zoned into three regions according to absorption strength: the white regions contain only structural aluminum and do not significantly contribute to absorption; the gray regions contain tantalum (Ta) as the primary absorber with a concentration of 30% by volume; and the black regions contain europium in the form of Eu{sub 2}O{sub 3} with a concentration of 33% by volume. A thick beryllium reflector is the final core component surrounding the CEs. A particular challenge to efficient maintenance and continued operation of HFIR is fabrication of the CEs, the current methods for which are tedious, outdated, and expensive. In an effort to improve the operational economics of HFIR and to demonstrate the application of advanced manufacturing techniques to the production of nuclear reactor components, the ultrasonic additive manufacturing (UAM) process is being investigated for fabrication of the HFIR CEs. This application of UAM involves directly embedding small cylindrical compacts (∼5.6 mm diameter, ∼1.7 mm thick) containing 60% neutron poison by volume into an aluminum matrix, using energy from ultrasonic vibrations to bond the compacts to the aluminum. The CEs contain discrete regions of absorbing material, in contrast to the standard design with homogeneously distributed absorbers. This novel design necessitates an investigation into how the HFIR core neutronic behavior differs from the case with homogeneous CEs so as to ascertain the feasibility of using additively manufactured CEs in HFIR. From an operations perspective, it is desirable for the discrete CEs to match the performance of the homogeneous CEs as closely as possible so as to avoid reconsideration of established safety margins. Earlier analyses of HFIR under beginning of cycle (BOC) conditions have indicated that although absorption properties on a local scale differ significantly between CE designs, the macroscopic behavior (neutron multiplication and flux and power distributions) of the HFIR core is not significantly altered. This study now addresses CE performance throughout depletion of the HFIR core during a reactor cycle. These analyses help to construct a more complete picture of the feasibility of employing additively manufactured CEs in HFIR and will reveal any adjustments which must be made to the design in order to maintain acceptable performance. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Combining RHF and HFIR Disposition Campaigns - Analysis, Opportunity, and Lessons Learned

Savannah River Nuclear Solutions has tasked its H Canyon facility to prepare for disposition of used Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) cores. HFIR cores consist of an inner and outer assembly of highly enriched uranium fuel plates. The two elements are annular, nest inside each other, and the fuel is a curved plate held between two collars. The H-Canyon facility has a dissolver insert specifically designed for HFIR cores and has performed this type of campaign before between 1976 and 1988. At H-Canyon, 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. The mission will be to dissolve a number of HFIR cores, purify the uranium, and down blend it to low enriched uranium suitable for commercial power reactors. HFIR has a sister core. A French high flux research reactor, called Reactor A Haut Flux (RHF), has nearly the same geometry as the outer element of HFIR. The annulus is only slightly smaller and the RHF active fuel height is {approx}97 cm while the HFIR active fuel height is only {approx}51 cm. SRNS is currently storing RHF cores (repatriated through circa 2003) awaiting a path for disposition. There is now an opportunity. RHF cores fit in the HFIR dissolver insert. Dispositioning them concurrent with HIFR avoids a future facility evolution of just to change dissolver inserts again. This work is the result of an initiative to propose appending the RHF cores to the anticipated HFIR campaign. Geometry and chemistry differences are evaluated and criticality safety assessed. The RHF cores were semi-explicitly modeled in various stages of dissolution. Normal and credible abnormal conditions were assessed. RHF was shown to be bounded by the dissolution of a complete HFIR core in both neutron multiplication and dissolver chemistry. Computational modeling was performed using KENOVI in the SCALE 6.1 code package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR

The High Flux Isotope Reactor (HFIR) is an ideal tool for materials irradiation testing because of its intense steady-state neutron flux. Many programs take advantage of HFIR’s central flux trap for irradiation experiments using capsules, also known as rabbits, to support advanced materials development and reactor design. The facility that makes up the HFIR flux trap has recently undergone a design change that increases the HFIR primary coolant volumetric flow rate by removing restrictions in the system. As a result, the usable cross-sectional area within the facility increased, opening the door to increase the cross-sectional area of the rabbit capsules that fill the facility. This report documents a new large-diameter rabbit housing that increases the usable volume within the rabbit capsule by 22.6%. However, challenges arise with increasing the capsule size, such as establishing a new maximum capsule operating pressure and determining the thermal-hydraulic characteristics. This report addresses those challenges with previously adopted HFIR safety methods. The rupture pressure of the rabbit housings is demonstrated while verifying that capsule swelling during and after rupture will not block coolant flow. Then, a safety factor is applied to ascertain an administrative operating pressure. Additionally, the thermal-hydraulic performance of the HFIR facility filled with large-diameter rabbit capsules is shown to not violate previously determined safety criteria. Next, heat transfer coefficients are determined for use in design calculations. Furthermore, this report gives an example of internal configurations for the new, larger rabbit capsules that use relevant boiling water reactor (BWR) cladding geometry. Finally, this report documents an example thermal design performance for a rabbit capsule containing six gauge-curved tensile tube specimens. The thermal performance gives predicted temperature distributions within the capsule and shows the expected temperature of the passive thermometers for post-irradiation temperature comparisons.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2025

Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Involute Working Group – Validation of CFD Turbulence Models for Steady-State Safety Analysis

There are three research reactors in the world having fuel plates curved as circle-involute (a spiral generated around a circle): The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) located in Tennessee, U.S.A. [ORNL, 2018]; The Laue-Langevin Institute (ILL) High Flux Reactor (RHF) located in Grenoble, France [ILL, 2018]; and, The Technical University of Munich (TUM) Research Neutron Source Heinz Maier-Leibnitz (FRM II) located in Garching, Germany [TUM, 2018]. All three reactors are currently using Highly Enriched Uranium (HEU, 235U/U ≥ 20 wt. %) as fuel and all three are actively engaged in activities to convert to Low Enriched Uranium (LEU, 235U/U < 20 wt. %) fuel. More information on these reactors can be found in Section two. For various reasons, these reactors have expressed interest in using Computational Fluid Dynamics (CFD) tools to perform their Steady-State Thermal-Hydraulic (SSTH) safety calculations. Using CFD tools for SSTH analysis represents generally a significant departure from traditional methods and requires thorough verification and validation (V&V) to be accepted by regulators. This is why Argonne National Laboratory (ANL) and the involute reactors formed an informal group to help each other in this endeavor. This so-called Involute Working Group (IWG) aims at supporting the qualification of CFD tools for SSTH safety calculations. Activities include benchmarking, code-to-code comparison, Verification and Validation (V&V) as well as technical support. More information on the IWG and the reasons why involute-plate reactors are interested in CFD can be found in Section three. The present study aims at testing three turbulence models commonly implemented in commercial software (namely K-Epsilon, K-Omega and Reynolds-Stress) and assessing their relevance for involute-plate reactor application. An experiment performed by Gambill and Bundy in 1961 was performed to measure heat transfer coefficients in conditions prototypic of the HFIR reactor [Gambill, 1961]. More details on this experiment are provided in Section four. Based on the information available, CFD models of this experiment have been created and the experimental results have been compared to the ones obtained with these CFD models. CFD models and the results of this comparison are provided in Section five.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Design Studies for the Optimization of 238 Pu Production in NpO 2 Targets Irradiated at the High Flux Isotope Reactor

Efforts to reestablish a domestic 238 Pu production capability in support of National Aeronautics and Space Administration mission objectives are ongoing throughout the U.S. Department of Energy complex. The Plutonium-238 Supply Project (PSP) was initiated in response to a report published by the National Research Council in 2011 stating that “without a restart of 238 Pu production, it will be impossible for the United States, or any other country, to conduct certain important types of planetary missions after this decade.” The PSP is targeting a sustained, constant production rate of 1.5 kg/year of heat source PuO 2 for several years. Design and optimization studies of 237 Np-bearing targets are underway at Oak Ridge National Laboratory (ORNL). It is anticipated that targets will be irradiated in ORNL’s High Flux Isotope Reactor (HFIR) and in the Advanced Test Reactor (ATR) at Idaho National Laboratory. A variety of target materials, containments, arrangements, and irradiation histories have been analyzed, and the results indicate that a sufficient quantity of 238 Pu can be produced in HFIR and ATR to fulfill the PSP’s constant production rate target. This paper focuses on the design and optimization of new target configurations containing pellets that are (1) ~93% of the theoretical density of NpO 2 , (2) loaded into pins of cladding materials that can be handled as solid waste following postirradiation 238 Pu recovery operations, (3) irradiated in various vertical experiment facility (VXF) locations in the HFIR permanent beryllium reflector, and (4) rotated within and/or moved to another VXF location following each HFIR operational cycle to maximize 238Pu production and minimize peak heat generation rates.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2025

Conceptual Designs for Irradiation Creep Testing of SiC in HFIR

Understanding irradiation creep of nuclear fuel cladding is important to properly size the initial fuel-cladding gap and understand when pellet-cladding contact is expected to occur due to a combination of fuel swelling and cladding creep-down. Irradiation creep also plays a role in relaxing stresses that develop in-pile. Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are the leading long-term accident-tolerant fuel cladding concept for light-water reactors (LWRs). Although some limited data are available regarding irradiation creep of the individual constituents (fibers, matrix), data regarding irradiation creep of SiC/SiC composites are currently insufficient. Additional data regarding irradiation creep compliance and the rupture lifetime (combination of creep and slow crack growth) are needed to understand material limitations. This work describes the design and development of two irradiation vehicles that are being pursued for testing SiC/SiC concepts in the High Flux Isotope Reactor (HFIR). The first is a passive experiment, referred to as the PRECISE experiment, that leverages the constant coolant pressure of HFIR to compress a metallic bellows and provide a well-characterized load to drive creep in a SiC/SiC dog bone specimen. The total creep strain would be quantified post-irradiation by measuring dimensional changes of the specimen length as well as local dimensional changes within the gauge region. Non-stressed specimens would also be irradiated under the same conditions to provide an indication of dimensional changes due to radiation-induced swelling in the absence of creep. A second, more complex experiment, referred to as the INSITE experiment, is being designed in parallel that would use pneumatics to pressurize a metal bellows and linear variable differential transformers (LVDTs) to measure the specimen displacement in situ during irradiation. Such an experiment would provide significantly more data regarding the evolution of the creep compliance as a function of dose and applied stress within a single experiment but would require significantly more development time and cost to execute. The primary concern with the INSITE experiment is the accuracy, reliability, and expected lifetime of the LVDTs during irradiation at elevated temperatures. Efforts are being made to adjust the experiment design and operating procedure to limit LVDT temperatures and mitigate or otherwise compensate for uncertainties due to factors such as temperature fluctuations, creep in the surrounding structural materials, and drift of the LVDTs. This work describes the experiment designs, thermal and structural analysis that were performed to ensure that the desired temperature and stress conditions can be achieved, some initial sensitivity analyses to predict the evolution of the radiation-induced specimen displacements, and potential sources of uncertainty in the measurements. Out-of-pile testing is being performed in parallel to confirm that the test trains achieve the expected stress states in the specimens and do not result in prohibitive stress concentrators (e.g., in the grip regions) that might risk pre-mature failure. The PRECISE experiments are proceeding toward fabrication and assembly with HFIR insertion planned during fiscal year 2026. The INSITE experiment is progressing toward out-of-pile demonstrations, which will provide more conclusive evidence regarding the feasibility of executing these tests in HFIR or whether alternative displacement monitoring techniques may need to be considered.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗