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This report reports the expansion of creep testing capabilities at ANL through the installation of 9 new creep frames in the Nuclear Science and Engineering Division. Data from the commissioning tests is presented.
We report given the current lack of long-term disposal for spent nuclear fuel storage in the United States, the current inventory of advanced test reactor (ATR) fuel at Idaho National Laboratory (INL) is expected to be stored in a dry storage facility for an extended storage period. As part of an effort to understand the conditions during a 50-year period, a three-dimensional computational fluid dynamics (CFD) of the unsealed canisters packed with spent fuel in the storage facility was constructed, this CFD model is coupled with radiolytic chemical reactions associated with water vapor and other species. The coupled model is simulated over a range of sensitivity parameters for the fuel decay heat, relative humidity, and oxyhydroxide film thickness for undried and fully-dried fuel. The most important parameter was the initial fuel decay heat. Over the range of cases, the hydrogen and nitric acid concentrations are low enough no significant impact is expected.
Advances in nuclear energy systems require an improved understanding of fuels, cladding, and structural materials; however, the extreme environmental conditions at irradiation test facilities often limit mechanical property evaluation to post-irradiation examination. In-situ monitoring techniques can accelerate materials qualification by providing direct insight into real-time material behavior during irradiation testing. Additively manufactured (AM) strain gauges offer a compact, material-compatible alternative to commercial sensors, particularly where space constraints, attachment limitations, or material compatibility requirements complicate conventional sensor use. This poster presents an overview of the development and testing of AM strain gauges in nuclear-relevant environments, demonstrating their potential to enhance structural health monitoring capabilities during irradiation experiments.
Isotope shifts of 223–226,228 Ra 19 F were measured for different vibrational levels in the electronic transition A 2 II 1/2 ←X 2 Σ + . The observed isotope shifts demonstrate the particularly high sensitivity of radium monofluoride to nuclear size effects, offering a stringent test of models describing the electronic density within the radium nucleus. Ab initio quantum chemical calculations are in excellent agreement with experimental observations. These results highlight some of the unique opportunities that short-lived molecules could offer in nuclear structure and in fundamental symmetry studies.
There is an imminent need for existing United States nuclear power plants to reduce their operating and maintenance costs to remain economically viable. Digital technology provides significant opportunity for the existing nuclear power plant fleet to transform that way in which work is accomplished to reduce costs and allow the fleet to remain economically competitive. However, a careful understanding of the human-technology integration is needed to ensure the continued safe and reliable operation of these existing plants with new digital capabilities. This work presents interim findings in applying human factors engineering to a safety-significant digital upgrade for a United States nuclear power plant, following the new Alternative Review Process in the recently revised Digital Instrumentation and Control Interim Staff Guidance Licensing Process, Revision 2. The interim results described in this work provides an industry perspective, based on ongoing work, to recent work published from Vazquez, Green, and Desaulniers (2022).
Here, the ring tension test (RTT) is a mechanical testing method for determining bulk mechanical behavior in the circumferential or hoop direction for tubular materials. The test is especially useful for testing materials with anisotropic mechanical properties, such as zirconium alloys, which are commonly used as nuclear fuel cladding. Anisotropy requires direction-specific testing to determine the hoop strength. Historically, several RTT methods and grips have been used, each method has its strengths and weaknesses, and, in all cases, the measured strength is subject to uncertainty due to variations of the testing geometry and experimental tolerances. Recent analysis has shown that grips with a hemicylindrical mandrel configuration are recommended as the most robust configuration. The two strictest aspects to be controlled are the ability to determine gage region orientation and closely matching the size of the mandrels to the test specimen. This last requirement is particularly challenging when the dimensions of the specimen vary because of environmental effects such as dimensional changes due to irradiation. This paper presents a new RTT grip designed to incorporate this mandrel shape, hold the gage at the desired orientation, be suitable for remote operation in a hot-cell environment, and be adaptable for different sizes or variations in the specimen size. The general description and the unique design features of the test specimen and grips are given in detail. The performance of the grips in mechanical testing, including in a remote hot-cell environment, is also provided.
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 .
In April 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) to advance US leadership in nuclear science and technology. By attracting new users from universities, laboratories, and industry, this program supports basic and applied nuclear research to help address the nation's energy security needs. In support of this program, the Idaho National Laboratory (INL) established in-house capabilities to develop, fabricate, test, and qualify new and enhanced temperature sensors for irradiation testing. This effort is continuing today through the DOE?s Advanced Sensors and Instrumentation (ASI) program. Although most efforts emphasize sensors capable of providing real-time data, selected tasks have been completed to enhance passive sensors for irradiations where instrumentation leads cannot be included. These sensors include silicon carbide (SiC) monitors, melt wires and the sublime temperature monitor. SiC monitors are available to detect peak irradiation temperatures between 200°C and 800°C in reactor locations where instrumentation leads cannot be used. SiC monitors may be evaluated using specialized equipment installed at INL?s Measurement Sciences Laboratory (MSL). A melt wire inventory is also maintained at MSL. This inventory contains wires for specific use in irradiation experiments ranging in temperatures from 30°C to 1500°C. Melt wires and SiC monitors have had decades of research and application. Recent research has produced a passive monitor known as the sublime temperature monitor. This passive sensor has the capability of recording temperature gradients. This paper will discuss passive temperature sensors currently being researched and implemented under the ASI program.
The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted to fill this gap. The two most used test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Finally, based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.
The process for developing and qualifying nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and postirradiation examinations; and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes ~20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In situ measurements generally provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the US nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the US Nuclear Regulatory Commission (NRC) outlined in its research information letter several technical issues that the industry must address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heat-up phase. By leveraging these advanced in situ capabilities, this work used in situ data generated from a simulated LOCA to better understand high-temperature creep and its effect on Zircaloy balloon and rupture performance. Here, this work used the BISON fuel performance code to assess the high-temperature creep model predictions with in situ data.
The present work offers and evaluates a correlation-based approach to determine the tensile properties of nuclear fuel cladding tubes. A set of model materials with known properties (i.e., austenitic and ferritic steels, fcc- and hcp-alloys) were tested to build a correlation between conventional tensile tests and ring (hoop) and tube (axial) tests. It was shown that close-to-linear relationships exist for the basic mechanical properties (i.e., yield and ultimate stress, uniform and total elongation) between the uniaxial tensile test data and the ring and tube test data. Using the ring and tube specimen geometries, the feasibility of the approach was demonstrated via manufacturing and testing specimens from irradiated Zr and FeCrAl nuclear fuel cladding tubes. Limitations caused by specific deformation mechanisms (e.g., high ductility of 304L steel) were discussed in detail. The approach can be extended to other tubular products with different geometry and dimensions.
This document outlines the Phase II test objectives and implementation plan for a down-select foam fixative technology intended to facilitate activities in support of the Savannah River Site (SRS) F/H labs deactivation and decommissioning (D&D) efforts. It is a collaborative effort between Savannah River National Laboratory (SRNL), Florida International University (FIU), and the SRS F/H labs team intended to test and evaluate the potential of a polyurethane resin foam in mitigating the release of contamination during dismantling operations on radioactively contaminated piping in legacy facilities. The cold test plan addresses specific requirements highlighted by site and safety personnel and will be executed at FIU and SRNL test and lab facilities. Results from the cold tests will inform the hot test at F/H labs, which will use the foam fixative to confine and/or isolate residual contamination within a 3-dimensional void space of Hastelloy C-22 piping designated for removal from the area and transported to a designated disposal facility. Phase I testing was previously conducted using Hilti CP620 fixative foam. Results from Phase I testing indicate Hilti CP-620 fixative foam is incompatible with SRS site hot taps as an effective foam delivery method into Hastelloy C-22 piping. Phase II testing will be conducted using FoamBag TM fixative foam as an alternative fixative foam option. Phase II testing will address eight test objectives: (1) evaluation of the adhesion and bonding properties of FoamBag TM in Hastelloy C-22 piping, (2) evaluation of the adhesion of FoamBag TM in piping under various moisture conditions, (3) determination of the heat profile of FoamBag TM during curing, (4) determination of the internal pipe pressure after FoamBag TM deployment and curing, (5) conduct a leak test to determine if FoamBag TM is effective at creating a full seal within piping, (6) headspace testing of FoamBag TM to determine if there are any associated off-gas hazards during FoamBag TM curing, (7) conduct environmental chamber testing of FoamBag TM to understand how environmental parameters impacts FoamBag TM curing, and (8) conduct fire testing to evaluate FoamBag TM fire retardant characteristics.
The Palisades Nuclear Generating Station included in its original surveillance program a surveillance capsule, designated A-60. The capsule was removed from its surveillance position in early 1995 and has been resident in the spent fuel pool since that time. It was harvested to perform characterization of surveillance specimens in this capsule in 2023. This capsule was irradiated to a fluence of 1.96x1020 n/cm2 (E> 1MeV) that is equivalent for more than 150 effective full power years for the current US reactor pressure vessel (RPV) fleet. This capsule contained several materials, including Charpy specimens of standard reference material (SRM) from Heavy-Section Steel Technology (HSST) A533-B Plate 01. To perform Charpy testing of this highly irradiated material, the new Charpy specimen transfer system was designed and implemented on Charpy impact testing machine in the hot cell to accommodate remote testing of these specimens. This new transfer system includes integrated environmental chamber to cool or heat Charpy specimens to the desired temperature. Testing of this highly irradiated material revealed very large shift of Charpy transition temperature, 188oC.
The development and verification of nuclear fuel-cladding performance codes require extensive testing to establish empirical correlations, necessitating more accelerated testing methods that can replace large validation studies. Here, in the present work, a technique was developed and implemented to experimentally quantify the relationship between temperature, cladding internal pressure, and strain in-situ during a burst test, with the specific aim of generating data for code validation and model refinement. Digital image correlation was used to measure hoop, axial, and radial strain, and infrared thermography to quantify axial temperature gradients. Several key experimental modifications to traditional LOCA burst testing were necessary, but are shown to effectively have little impact on burst temperatures. The time/temperature dependent pressure and time/spatially dependent thermal gradient data were used with BISON to simulate cladding burst and strain rates. Although the measured DIC strains reach appreciable amounts at slightly lower temperatures than BISON simulated strains, there is good agreement between the measured and simulated strains in terms of magnitude and rates leading up to burst. However, in the few seconds prior to and during burst, the BISON simulated strains are significantly lower than the measured strains, indicating the potential for model improvement. The combined experiment and simulation technique may be applied to any new cladding concept to accelerate fuel qualification.
To ensure the safe disposal of radioiodine-containing nuclear wastes, a durable iodine waste form (IWF) is required. Many IWFs have been developed and evaluated for chemical durability using a wide range of test methods, making direct comparisons difficult as highlighted in this work through a review of available IWF corrosion data. To move toward a common dataset, a semi-dynamic leach test suite was experimentally demonstrated using several IWF types, providing standard comparisons. The results highlighted clear differences in corrosion resistance between various IWFs. This test suite is suggested for future studies to generate comparable data for a common dataset.
The contribution of transient fission gas release (tFGR) to the burst behavior of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant additional transient fission gas release in excess of the steady state fission gas release from normal operation. An addition to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) was created to explore the different conditions that impact transient fission gas release. Four tests on commercially irradiated zirconium clad uranium dioxide nuclear fuel were performed to support transient fission gas release study. These experiments and their post-test analyses are discussed. The first test on high burnup fuel revealed a LOCA relevant transient fission gas release of 5.3% for an unpressurised segment. Subsequent tests have measured transient fission gas release at different conditions that span from 4.8% to 14.4 % transient fission gas release.
The aim for storage of spent nuclear fuel (SNF) either in wet or in dry storage systems is to ensure general safety objective s are met throughout a desired storage period. Staff at the Savannah River National Laboratory (SRNL), in collaborations with partners at other national laboratories, industry research organizations, and the University of South Carolina (UofSC), have performed materials aging testing and analyses, and have established nuclear materials aging management programs to support extended periods of safe storage of research reactor (RR) SNF and of commercial power reactor (PR) SNF pending ultimate disposal. Several example challenges include susceptibility of aluminum SNF from research reactors to corrosion in poor quality water (wet storage). In dry storage, aluminum SNF can release hydrogen via radiolysis of the hydrated oxides on the aluminum cladding. Austenitic stainless steel canisters used for dry storage are susceptible to chloride-induced stress corrosion cracking (outside-in attack) that threaten the confinement boundary provided by the canister. This paper further describes these challenges, among others, and the formulated solutions to support extended safe storage of SNF.