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DOE OSTI2017

Development and implementation of an in-situ cable condition monitoring method for nuclear power plants - 130

The sustainability of the existing fleet of nuclear power reactors depends on utilities' ability to manage the aging of key components that cannot easily or economically be replaced or upgraded. Cables, metals, and concrete structures are among the major and key components that the industry and government are working on to ensure the long term reliability and safety of nuclear plants. To ensure that these key components continue to meet their performance requirements, new testing and condition monitoring technologies must be developed that are capable of identifying and quantifying degradation in these materials. The work presented in this paper focuses on the development and implementation of an in-situ condition monitoring method for nuclear power plant cables. This cable condition monitoring method is based on the frequency domain reflectometry (FDR) technique, which is a non-destructive in-situ electrical test that uses the principle of transmission line theory to locate and quantify impedance changes in a cable circuit. Through extensive research, development, and validation, the FDR test technique has been adapted as a cable condition monitoring method that trends with increasing cable degradation. In addition, the FDR data has been correlated to the industry standard elongation at break (EAB) test to quantify the degree of localized degradation in nuclear plant cable insulation material caused by long term exposure to high temperatures and radiation. By correlating FDR to EAB, the aging condition of cables routed through harsh environments can be better understood in relation to its ability to perform its intended function. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2011

Status of Initial Assessment of Physical and Mechanical Properties of Graphite Grades for NGNP Appkications

Current candidate graphite grades for the core structures of NGNP include grades NBG-17, NBG-18, PCEA and IG-430. Both NBG-17 and NBG-18 are manufactured using pitch coke, and are vibrationally molded. These medium grain products are produced by SGL Carbon SAS (France). Tayo Tanso (Japan) produces IG-430 which is a petroleum coke, isostatically molded, nuclear grade graphite. And PCEA is a medium grain, extruded graphite produced by UCAR Carbon Co. (USA) from petroleum coke. An experimental program has been initiated to develop physical and mechanical properties data for these current candidate graphites. The results will be judged against the requirements for nuclear grade graphites set forth in ASTM standard D 7219-05 "Standard Specification for Isotropic and Near-isotropic Nuclear Graphites". Physical properties data including thermal conductivity and coefficient of thermal expansion, and mechanical properties data including tensile, compressive and flexural strengths will be obtained using the established test methods covered in D-7219 and ASTM C 781-02 "Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors". Various factors known to effect the properties of graphites will be investigated. These include specimen size, spatial location within a graphite billet, specimen orientation (ag and wg) within a billet, and billet-to-billet variations. The current status of the materials characterization program is reported herein. To date billets of the four graphite grades have been procured, and detailed cut up plans for obtaining the various specimens have been prepared. Particular attention has been given to the traceability of each specimen to its spatial location and orientation within a billet.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2024

Sister Rod Destructive Examinations (FY23)

As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report provides a cumulative status report of the testing progress, data accumulated, and conclusions reached as a part of the program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Blind Modeling Validation Exercises Using the Horizontal Dry Cask Simulator

The U.S. Department of Energy (DOE) established a need to understand the thermal-hydraulic properties of dry storage systems for commercial spent nuclear fuel (SNF) in response to a shift towards the storage of high-burnup (HBU) fuel (> 45 gigawatt days per metric ton of uranium, or GWd/MTU). This shift raises concerns regarding cladding integrity, which faces increased risk at the higher temperatures within spent fuel assemblies present within HBU fuel compared to low-burnup fuel (≤ 45 GWd/MTU). A dry cask simulator (DCS) was built at Sandia National Laboratories (SNL) in Albuquerque, New Mexico to produce validation-quality data that can be used to test the accuracy of the modeling used to predict cladding temperatures. These temperatures are critical to evaluating cladding integrity throughout the storage cycle of commercial spent nuclear fuel. A model validation exercise was previously carried out for the DCS in a vertical configuration. Lessons learned during the previous validation exercise have been applied to a new, blind study using a horizontal dry cask simulator (HDCS). Three modeling institutions – the Nuclear Regulatory Commission (NRC), Pacific Northwest National Laboratory (PNNL), and Empresa Nacional del Uranio, S.A., S.M.E. (ENUSA) – were granted access to the input parameters from the DCS Handbook, SAND2017-13058R, and results from a limited data set from the horizontal BWR dry cask simulator tests reported in the HDCS update report, SAND2019-11688R. With this information, each institution was tasked to calculate peak cladding temperatures and air mass flow rates for ten HDCS test cases. Axial as well as vertical and horizontal transverse temperature profiles were also calculated. These calculations were done using modeling codes (ANSYS/Fluent, STAR-CCM+, or COBRA-SFS), each with their own unique combination of modeling assumptions and boundary conditions. For this validation study, the ten test cases of the horizontal dry cask simulator were defined by three independent variables – fuel assembly decay heat (0.5 kW, 1 kW, 2.5 W, and 5 kW), internal backfill pressure (100 kPa and 800 kPa), and backfill gas (helium and air). The plots provided in Chapter 3 of this report show the axial, vertical, and horizontal temperature profiles obtained from the dry cask simulator experiments in the horizontal configuration and the corresponding models used to describe the thermal-hydraulic behavior of this system. The tables provided in Chapter 3 illustrate the closeness of fit of the model data to the experiment data through root mean square (RMS) calculations of the error in peak cladding temperatures (PCTs), PCT axial locations, axial temperature profiles, vertical and horizontal temperature profiles at two different axial locations, and air mass flow rates for the ten test cases, normalized by the experimental results. The model results are assigned arbitrary model numbers to retain anonymity. Due to the relatively flat axial temperature profiles, small temperature gradients resulted in large deviations of all models’ PCT axial location from the experimental PCT axial location. When the PCT axial location error is excluded in the calculation of the combined RMS of the normalized errors that considers PCT, the temperature profiles, and the air mass flow rates, the model data fits the experimental data to within 5%. When the vault information is excluded, the model data fits the experimental data to within 2.5%. An error analysis was developed further for one model, using the model and experimental uncertainties in each validation parameter to calculate validation uncertainties. The uncertainties for each parameter were used to define quantifiable validation criteria. For this analysis, the model was considered validated for a given comparison metric if the normalized error in that metric divided by the validation uncertainty was less than or equal to 1. When considering the combined RMS of the normalized errors of all metrics divided by their validation uncertainties, the model was found to have satisfied the criterion for model validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Review of Data from Shock Physics Experiments SPE-1, SPE-2 and SPE-3: Data Corrections and Phenomenology Analysis

The NNSA is conducting a series of high explosive (HE) tests – the Source Physics Experiments – at the National Nuclear Security Site (NNSS). These experiments are intended to investigate the generation of geophysical signals from controlled sources. These signals will aid in the development of physics-based models to aid in the discrimination of signals emitted from these and other possible sources, such as earthquakes and clandestine nuclear events. The SPE events include an array of accelerometers near to the source intended to provide measurements of the strong ground motion, or “near field,” regime. LANL performed an extensive review of the near field measurements from the first two events (SPE-1 and SPE-2) in this series. This review revealed that the near field gauges likely rotated during installation, causing measurement uncertainty. 2 This conclusion led to the installation of new gauges for the third experiment, SPE-3, utilizing new physical design and improved installation practices to achieve better quality control on their orientation. A subsequent analysis of SPE-3 data confirmed that rotations occurred and provided the basis for corrections to those data where the amount of rotation could be credibly determined. The corrected data supports a more confident analysis of the entire near field data set. This report summarizes the data errors and provides results of the analysis of the corrected dataset.

42 ENGINEERING↗

DOE OSTI2013

A modified detector concept for SuperCDMS: The HiZIP and its charge performance

SuperCDMS (Super Cryogenic Dark Matter Search) is a leading direct dark mat-ter search experiment which uses solid state detectors (Ge crystals) at milliKelvintemperatures to look for nuclear recoils caused by dark matter interactions in the de-tector. `Weakly Interacting Massive Particles' (WIMPs) are the most favoured darkmatter candidate particles. SuperCDMS, like many other direct dark matter searchexperiments, primarily looks for WIMPs. The measurement of both the ionizationand the lattice vibration (phonon) signals from an interaction in the detector allow itto discriminate against electron recoils which are the main source of background forWIMP detection.SuperCDMS currently operates about 9 kg of Ge detectors at the Soudan under-ground lab in northern Minnesota. In its next phase, SuperCDMS SNOLAB plansto use 100-200 kg of target mass (Ge) which would allow it to probe more of theinteresting and and as of yet unexplored parameter space for WIMPs predicted bytheoretical models. The SuperCDMS Queen's Test Facility is a detector test facilitywhich is intended to serve as detector testing and detector research and developmentpurposes for the SuperCDMS experiment.A modifed detector called the HiZIP (Half-iZIP), which is reduced in complex-ity in comparison to the currently used iZIP (interleaved Z-sensitive Ionization and Phonon mediated) detectors, is studied in this thesis. The HiZIP detector designalso serves to discriminate against background from multiple scatter events occurringclose to the surfaces in a single detector. Studies carried out to compare the surfaceevent leakage in the HiZIP detector using limited information from iZIP data takenat SuperCDMS test facility at UC Berkley produce a highly conservative upper limitof 5 out of 10,000 events at 90% condence level. This upper limit is the best amongmany different HiZIP congurations that were investigated and is comparable to theupper limit calculated for an HiZIP detector in the same way using the same data. Areal HiZIP device operated at the Queen's Test Facility produced an exposure lim-ited 90% upper limit of about 1 in 100 events for surface event leakage. The dataused in these studies contain true nuclear recoil events from cosmogenic and ambientneutrons. This background was not subtracted in the calculation of the upper limitsstated above and hence they are highly conservative.A surface event source was produced by depositing210Pbfrom radon exposureonto a copper plate. This source was then used to take data for a surface eventdiscrimination study of the HiZIP detector operated at the Queen's Test Facility.A study of the contribution of the noise from capacitive crosstalk between chargesensors in a HiZIP detector configuration was investigated, conrming the expectationthat no signicant drop in performance compared to an HiZIP is to be expected dueto this effect.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2016

A comparison of the irradiation creep behavior of several graphites - 18583

Graphite creep strain data from the irradiation creep capsule Advanced Graphite Creep-1 (AGC-1) are reported. This capsule was the first (prototype) of a series of five or six capsules planned as part of the AGC experiment, which was designed to fully characterize the effects of neutron irradiation and the radiation creep behavior of current nuclear graphite. The creep strain data and analysis are reported for the six graphite grades incorporated in the capsule. The AGC-1 capsule was irradiated in the Advanced Test Reactor at Idaho National Laboratory (INL) at approximately 700 deg. C and to a peak dose of 7 dpa (displacements per atom). The specimen s final dose, temperature, and stress conditions have been reported by INL and were used during this analysis. The derived creep coefficients (K) were calculated for each grade and were found to compare well to literature data for the creep coefficient, even under the wide range of AGC-1 specimen temperatures. Comparisons were made between AGC-1 data and historical grade data for creep coefficients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Evaluation of Radiolysis Data for Hydrogen Gas Generation During Gamma Irradiation of Pre-Corroded and Pristine Aluminum Samples - An Aluminum SNF Dry Storage Study Interim Report

Information and data from radiolysis testing to measure hydrogen (H₂) generated from hydrated oxides on aluminum exposed to ⁶⁰Co radiation were reviewed to evaluate hydrogen generation with radiation dose. Hydrogen generation rate is a primary input to the coupled thermal-chemical system model of the performance of aluminum-clad spent nuclear fuel in a dry storage canister (ASNF-in-canister) over its storage life. Hydrogen generation data and hydrogen generation rate (hydrogen generated per absorbed radiation dose, $\Delta$(H₂)/$\Delta$(dose)) were reported in two reports for radiolysis testing [1, 2]. The testing involved first-time data¹ for radiolytic yield of hydrogen from hydrated-oxides-on-aluminum substrates, and from pristine (non-corroded) aluminum substrates. The laboratory-grown hydrated-oxides-on-aluminum substrates contained a ~ 5 μm film consisting of the trihydroxide bayerite (Al(OH)₃) with amounts of the oxyhydroxide boehmite (AlOOH). These specimens were assumed to also contain physisorbed water (unquantified). The pristine substrates were flat coupons of aluminum that were ground to a 600-grit finish but were not immersed in water to grow a hydrated oxide. The pristine substrates were tested as companion specimens to provide information on hydrogen generation from material without chemisorbed water and were assumed to contain physisorbed water (unquantified).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

The Mechanical Response of High Burnup 17x17 PWR Fuel Rods Under Bending

As a part of the US Department of Energy Office of Nuclear Energy High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory has performed 4pt-bend tests on a series of 17x17 PWR fuel rods that were discharged with rod average burnups between 50 and 60 GWd/MTU. The rods include four different kinds of fuel cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Six rods of each type (combining Zirc-4 and LT Zirc-4 as one type) were tested to failure at room temperature and 200 oC. Three of each type experienced a full-length heat treatment cycle from RT to 400 oC to RT with an 8 hr soak at 400 oC to simulate the expected limiting vacuum drying condition, which potentially induces hydride reorientation and causes some annealing of irradiation damage in the cladding. Overall, the fuel rods show significant capacity to deflect in bending and exhibit non-linear deformation before failing. The calculated failure strains range from 1.5% to 3.8%. The simulated vacuum drying heat treatment had the largest effect on fuel rods with ZIRLO and M5 cladding. In both cases the heat treatment caused increases in the plastic strain at failure, suggesting some annealing of irradiation damage is occurring. The finite element modeling of the 4pt bending test assume the pellet is a collection of pellet wedge pieces. The results for the Zircaloy-4 rods suggest that the pellets are bonded to the cladding and friction controls the relative motion of the pellet wedge pieces.

Cantonwine, Paul↗

DOE OSTI2023

On Magnitudes and Inferred Yields of the 39 Underground Nuclear Test Explosions at the Novaya Zemlya Test Site

ABSTRACT We review the previously compiled seismic magnitudes for underground nuclear explosions at the Union of the Soviet Socialist Republics Novaya Zemlya Test Site, and analyze these data in the context of published details about individual tests and official Russian reports of annual cumulative explosive energy release at the test site. We derive relationships between teleseismic body wave magnitude mb and explosive yield Y by predicting the yield of each test using mb and minimizing the difference between predicted and reported annual cumulative yield. A linear relationship between mb and logY with a slope of 0.82 provides a good fit to the annual total yield, with a root mean square (rms) misfit of 0.095 logY units. A spectral magnitude relationship based on the explosion source model of Mueller and Murphy (1971) with a single-free-parameter provides a slightly better fit. For the 10 tests for which individual yields can be directly deduced from the official reports, the rms misfit of reported yields is less than 0.1 logY units or about 25%. More complex relationships that differentiate between the northern and southern test areas, or take account of variations in the scaled depth of burial, do not provide substantially better fits to the reported yields.

Geochemistry & Geophysics↗

DOE OSTI2023

A Review of the Infrasonic Yield-Period Relation and Some Additional Comments on Observed Azimuth Deviations

In this report we will review the empirically derived infrasonic yield-period relation for atmospheric nuclear explosions. Data came from infrasonic arrays operated by the US Government during the period of atmospheric testing. Only US tests with announced yields will be discussed, and data presented here does not cover every US test because the observed periods came from a set of old Air Force Technical Applications Center (AFTAC) technical memos (TM). The TMs reported the average periods used to obtain a yield-period relation and we use those periods here; the period data from three of the Test Operations are in declassified reports, (see WT reports in References). In addition our analysis employs the announced unclassified yields.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2023

Scaling Containment on a Large Centrifuge

Containment science concerns the trapping or leakage of radioactive cavity gases from underground nuclear explosions (UNEs). The most physically relevant data for validation of containment science comes from legacy UNE testing or large field-scale chemical explosions. Field-scale tests can inform on the degree of residual stress imparted to host geology and possible formation of an accompanying gas-containing “stress cage”, for example, as a function of chemical explosion yield. However, field testing can be expensive, difficult to conduct, and challenging to instrument with limitations on data coverage (i.e., the number of boreholes; difficult coring through damage zones). Small-scale laboratory tests are typically simpler to conduct and more thoroughly characterize, but may not include the right scale of containment processes for stress-cage or chimney formation. This report presents the theory and fundamentals of large centrifuge physical modeling, including a history of explosive or similar testing relevant to containment science investigations on centrifuges. Subscale models with embedded chemical explosives in the enhanced gravity of a large centrifuge can represent hundreds of meters of depth and large explosions not otherwise attainable in the laboratory as based on scaling of length, energy, and other processes by the g-factor of the centrifuge (i.e., the number of times larger the centrifugal force is than the gravitational force on the Earth’s surface). The centrifuge uniquely joins the simplicity and exhaustiveness of laboratory-scale characterization with the physics of field-scale processes. This report is part of Sandia National Laboratories’ (SNL’s) effort to prepare for physical modeling with approximately one-meter tall geologic models of stress cage and/or chimney formation on its 29-foot [8.84 m] radius centrifuge with a load capacity of 1.6 million 𝑔-pounds [726 𝑔-ton]. The scaling relationships herein will inform model design decisions and performance requirements for in-flight sensors.

42 ENGINEERING↗

DOE OSTI2013

Radiation Embrittlement Archive Project

The Radiation Embrittlement Archive Project (REAP), which is being conducted by the Probabilistic Integrity Safety Assessment (PISA) Program at Oak Ridge National Laboratory under funding from the U.S. Nuclear Regulatory Commission s (NRC) Office of Nuclear Regulatory Research, aims to provide an archival source of information about the effect of neutron radiation on the properties of reactor pressure vessel (RPV) steels. Specifically, this project is an effort to create an Internet-accessible RPV steel embrittlement database. The project s website, https://reap.ornl.gov, provides information in two forms: (1) a document archive with surveillance capsule(s) reports and related technical reports, in PDF format, for the 104 commercial nuclear power plants (NPPs) in the United States, with similar reports from other countries; and (2) a relational database archive with detailed information extracted from the reports. The REAP project focuses on data collected from surveillance capsule programs for light-water moderated, nuclear power reactor vessels operated in the United States, including data on Charpy V-notch energy testing results, tensile properties, composition, exposure temperatures, neutron flux (rate of irradiation damage), and fluence, (Fast Neutron Fluence a cumulative measure of irradiation for E>1 MeV). Additionally, REAP contains data from surveillance programs conducted in other countries. REAP is presently being extended to focus on embrittlement data analysis, as well. This paper summarizes the current status of the REAP database and highlights opportunities to access the data and to participate in the project.

Klasky, Hilda B↗

DOE OSTI2015

Benchmark Testing of a New 56 Fe Evaluation for Criticality Safety Applications

The SAMMY code was used to evaluate resonance parameters of the 56 Fe cross section in the resolved resonance energy range of 0–2 MeV using transmission data, capture, elastic, inelastic, and double differential elastic cross sections. The resonance analysis was performed with the code SAMMY that fits R-matrix resonance parameters using the generalized least-squares technique (Bayes’ theory). The evaluation yielded a set of resonance parameters that reproduced the experimental data very well, along with a resonance parameter covariance matrix for data uncertainty calculations. Benchmark tests were conducted to assess the evaluation performance in benchmark calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Status Report for NEET In-Pile Sensor Irradiation Capabilities at TREAT

The recently-established in-pile sensor development project under the Nuclear Energy Enabling Technology (NEET) program is focused on the development of measurement technologies to enable acquisition of data focused on the performance, properties, and boundary conditions of nuclear fuels/materials under irradiation in nuclear reactors. These sensors can be used in steady state irradiation tests to measure parameters of interest under conditions representing normal nuclear plant operation, or in transient irradiation tests to determine specimen performance under postulated unplanned conditions. In either case, development of suitable sensor technologies must include instrument irradiations to help guide sensor material/design selection, verify instrument performance, and demonstrate deployment viability. The Transient Reactor Test facility (TREAT) while just a part of the overall capabilities portfolio needed to enable this sensor development mission, has unique capabilities and characteristics that make it well poised for a number of instrument irradiation data objectives including sensor survivability screening, response characterization, qualification tests, and in-situ properties measurements. TREAT’s brilliantly-basic design, when married with appropriate irradiation hardware and analytic capabilities, fills a crucial capability gap between out-of-pile testing and irradiation in steady state high power research reactors. Leveraging a recently developed approach to irradiation testing in TREAT, the present work demonstrates that cost-effective and scientifically-impactful instrument irradiation capabilities for TREAT are a well-placed capability in the NEET in-pile sensor development project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Enhancements in SCALE 6.1

The SCALE code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a 'plug-and-play' framework with 89 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.1 builds on the existing capabilities and ease-of-use of SCALE and provides several new features such as enhanced lattice physics capabilities and multigroup Monte Carlo depletion, improved options and capabilities for sensitivity and uncertainty analysis calculations, improved flexibility in shielding and criticality accident alarm system calculations with automated variance reduction, and new options for the definition of group structures for depletion calculations. The SCALE 6.1 development team has focused on improved robustness via substantial additional regression testing and verification for new and existing features, providing improved performance relative to SCALE 6.0, especially in reactor physics calculations and in the nuclear data used for source term characterization and shielding calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Silicon Carbide Temperature Monitor Processing Improvements. Status Report

Silicon carbide (SiC) temperature monitors are used as temperature sensors in Advanced Test Reactor (ATR) irradiations at the Idaho National Laboratory (INL). Although thermocouples are typically used to provide real-time temperature indication in instrumented lead tests, other indicators, such as melt wires, are also often included in such tests as an independent technique of detecting peak temperatures incurred during irradiation. In addition, less expensive static capsule tests, which have no leads attached for real-time data transmission, often rely on melt wires as a post-irradiation technique for peak temperature indication. Melt wires are limited in that they can only detect whether a single temperature is or is not exceeded. SiC monitors are advantageous because a single monitor can be used to detect for a range of temperatures that occurred during irradiation. As part of the process initiated to make SiC temperature monitors available at the ATR, post-irradiation evaluations of these monitors have been previously completed at the High Temperature Test Laboratory (HTTL). INL selected the resistance measurement approach for determining irradiation temperature from SiC temperature monitors because it is considered to be the most accurate measurement. The current process involves the repeated annealing of the SiC monitors at incrementally increasing temperature, with resistivity measurements made between annealing steps. The process is time consuming and requires the nearly constant attention of a trained staff member. In addition to the expensive and lengthy post analysis required, the current process adds many potential sources of error in the measurement, as the sensor must be repeatedly moved from furnace to test fixture. This time-consuming post irradiation analysis is a significant portion of the total cost of using these otherwise inexpensive sensors. An additional consideration of this research is that, if the SiC post processing can be automated, it could be performed in an MFC hot cell, further reducing the time and expense of lengthy decontaminations prior to processing. Sections of this report provide a general description of resistivity techniques currently used to infer peak irradiation temperature from silicon carbide temperature monitors along with some representative data, the proposed concepts to improve the process of analyzing irradiated SiC temperature monitors, the completed efforts to prove the proposed concepts, and future activities. The efforts detailed here succeeded in designing and developing a real-time automated SiC resistivity measurement system, and performed two initial test runs. Activities carried out include the assembly and integration of the system hardware; the design and development of a preliminary monitor fixture; the design of a technique to automate the data analysis and processing; the development of the communication, coordination, and user software; and the execution and troubleshooting of test run experiments using the box furnace. Although the automation system performed as required, the designed fixture did not succeed in establishing the needed electrical contacts with the SiC monitor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

NDMAS

Overview of Current ART-GCR Data: Fuel Fabrication, Irradiation Monitoring (Fuel & Graphite – near real-time for HDG-1), Post-Irradiation Examination (Fuel & Graphite), Graphite Characterization (Baseline and Irradiated), High Temperature Metals Mechanical Tests, Design, Methods, and Validation Data, Japan Atomic Energy Agency’s High Temperature Test Reactor (HTTR), Argonne National Laboratory’s Natural convection Shutdown heat removal Test Facility (NSTF), Oregon State University’s High Temperature Test Facility (HTTF), Generation IV International VHTR Materials Handbook, Additional related data, and Advanced Test Reactor operations (near real-time).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗