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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

EVALUATION OF PRESSURE DROP CORRELATIONS FOR THE WIRE-WRAPPED ROD BUNDLES

The accurate prediction of pressure drop in the fuel rod bundle is of importance for the design and thermal-hydraulic analysis of nuclear reactor system. Due to very complex geometry of wire-wrapped fuel rod bundle geometry for the Sodium Fast Reactor (SFR), a simple method using hydraulic (equivalent) diameter is not sufficient to predict the pressure drop of rod bundle accurately. For several decades, many experimental works have been performed to develop an accurate correlation for wire-wrapped fuel bundle friction factor. In these works, the friction factor correlation of wire-wrapped fuel bundle has been developed as a function of Reynolds number of the flow and the design parameters such as fuel pin diameter, wire diameter, rod pitch, and wire lead length. The Cheng and Todreas correlations are well-known and used in many subchannel analysis code, such as CADET, TRIO-U, MATRA-LMW, SE2-ANL, etc. Novendstern model, Rehme model, Baxi and Dalle Donne model are also well-known and widely referred. Computational Fluid Dynamics is another typical approach to predict the pressure drop of wire-wrapped fuel rod bundle, but it requires the model validation because the accuracy of pressure drop prediction by the CFD simulation depends on what turbulence model is employed. The present work aims at evaluating the existing pressure drop correlations for the wire-wrapped fuel rod bundle and evaluating the pressure drop of 217-pin rod bundle by the CFD analysis.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Determination of water equivalent content in fuel rods

In water-cooled reactors, hydriding failure of zircaloy cladding tubes can be easily caused by high content of total water in fuel rods. WEC, which is the water equivalent content within the cold space of a fuel rod, directly reflects the total water content of the fuel rod. A device measuring WEC has been designed, manufactured and tested. This extraction device is a hermetically sealed system consisting of six parts: the break-off part, the heating unit, the reduction unit, the sample storage part, the collecting unit and the gas flow system. Each part is connected to another one using a special connection, and the device's end part is connected to a gas chromatograph. The determination procedure can be briefly summarized as follows: first, a test fuel rod should be put into the sealed device after lathing two circular grooves on both ends, so that it can be easily broken at these points in the break-off part. Driven by carrier gas, water and hydrogen both contained in helium gas and absorbed on the rod in-wall pass over the hot uranium filament where water is converted into hydrogen. After being concentrated by the collecting unit, hydrogen flows into the gas chromatograph for measurement and data are calculated as total water content A. UO{sub 2} pellets in the test rod should be taken out from the sample storage part and the total hydrogen content of these pellets is determined according to ASTM C1457 standard, followed by its conversion into total water content B. Finally, the sum of A and B being divided by the free volume of the cold space within the test rod, gives the value of WEC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Acoustic sensors devoted to the detection of leaking irradiated fuel rod in spent fuel pool

The acoustic sensor has been specially designed in order to provide a non-destructive technique to easily carry out the measurement of the internal gas pressure and gas composition of a LWR nuclear fuel rod. This sensor has been achieved in 2007 and is now covered by an international patent. In 2015, we presented a generation of acoustic sensors devoted to the characterization of fission gas. The gas composition can be defined with an uncertainty of about 1% and the pressure with an uncertainty of about 10 % (for around 50 bars and a gas mixture containing 20% Xe/Kr). This measurement is possible in hot laboratory facility using viscous liquid to couple the sensor with the rod. We develop a new acoustic sensor for an application in a spent fuel pool without viscous liquid to couple the sensor with the rod. This application consists in determining the default rod in a fuel assembly which has been declared 'no leak-proof' during the fuel discharging of the reactor. We develop a new method based on the detection of gas pressure drop into the fuel rod. This new method could replace the current one which is based on the detection of water into the fuel rod. The new generation of transducer and the associated methodology are now operational in laboratory for nondestructive measurement under water and enable to detect the gas pressure drop into the fuel rod. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Simulation of Proposed Heat Treatment Methods for the Sister Rods

The High Burnup Spent Fuel Data Project (the 'project'), sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term dry storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for analytical model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask (the Research Project Cask or RPC) at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (Project Assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO{sup R}, and M5{sup R}. In approximately ten years, the instrumented cask will be opened for inspection and several fuel rods from the Project Assemblies will be characterized. The project has also removed 25 'sister rods' from assemblies at North Anna: 9 directly from the project assemblies and 16 from similarly constructed and operated HBU assemblies. The 25 sister rods are all high burnup and cover the range of domestically-used cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive examinations (NDE) followed by destructive examination (DE). Examination of the sister rods will provide a baseline for comparison with the cask rods and will allow for separate effects testing. Dry storage canister loading is completed under water in a designated area of the plant's spent fuel pool. In preparation for dry storage, the fuel assemblies and canister cavity must be drained and dried. Typically, the highest temperatures experienced by the fuel during dry storage are encountered either during drying or transfer of the canister to the storage pad; therefore, as a part of the sister rod testing these temperatures will be simulated in the lab to produce samples that are representative of the peak cladding temperature the fuel experiences either during drying, transfer, or storage. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Growth and characterizations of GaN micro-rods on graphene films for flexible light emitting diodes

We report the growth of GaN micro-rods and coaxial quantum-well heterostructures on graphene films, together with structural and optical characterization, for applications in flexible optical devices. Graphene films were grown on Cu foil by means of chemical vapor deposition, and used as the substrates for the growth of the GaN micro-rods, which were subsequently transferred onto SiO{sub 2}/Si substrates. Highly Si-doped, n-type GaN micro-rods were grown on the graphene films using metal–organic chemical vapor deposition. The growth and vertical alignment of the GaN micro-rods, which is a critical factor for the fabrication of high-performance light-emitting diodes (LEDs), were characterized using electron microscopy and X-ray diffraction. The GaN micro-rods exhibited promising photoluminescence characteristics for optoelectronic device applications, including room-temperature stimulated emission. To fabricate flexible LEDs, In{sub x}Ga{sub 1–x}N/GaN multiple quantum wells and a p-type GaN layer were deposited coaxially on the GaN micro-rods, and transferred onto Ag-coated polymer substrates using lift-off. Ti/Au and Ni/Au metal layers were formed to provide electrical contacts to the n-type and p-type GaN regions, respectively. The micro-rod LEDs exhibited intense emission of visible light, even after transfer onto the flexible polymer substrate, and reliable operation was achieved following numerous cycles of mechanical deformation.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Parallel Operation of Multiple Closely Spaced Small Aspect Ratio Rod Pinches

A series of simulations and experiments to resolve questions about the operation of arrays of closely spaced small aspect ratio rod pinches has been performed. Design and post-shot analysis of the experimental results are supported by 3D particle-in-cell simulations. Both simulations and experiments support these conclusions. Penetration of current to the interior of the array appears to be efficient, as the current on the center rods is essentially equal to the current on the outer rods. Current loss in the feed due to the formation of magnetic nulls was avoided in these experiments by design of the feed surface of the cathode and control of the gap to keep the electric fields on the cathode below the emission threshold. Some asymmetry in the electron flow to the rod was observed, but the flow appeared to symmetrize as it reached the end of the rod. Interaction between the rod pinches can be controlled to allow the stable and consistent operation of arrays of rod pinches.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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 OSTI2017

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Sister Rod Nondestructive Examination Final Report

This report documents work performed under the Spent Fuel and Waste Disposition for the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Science and Technology program. This work was performed to fulfill the Level 2 Milestone M2SF-17OR010201021, “Documentation of Non-destructive Tests on Sister Pins,” within work package SF-17OR01020102. The High Burnup Spent Fuel Data Project, sponsored by DOE-NE, is focused on understanding the effects of long-term storage and transportation on high burnup (HBU) (>45 gigawatt days per metric ton uranium) light water reactor fuel. The goals of the project are to “provide confirmatory data for model validation and potential improvement, provide input to future spent nuclear fuel (SNF) dry storage cask design, support license renewals and new licenses for Independent Spent Fuel Storage Installations (ISFSIs), and support transportation licensing for high burnup SNF". In support of project goals, 25 sister rods were removed from fuel assemblies at the North Anna Power Station. Nine rods were removed from the project assemblies, and 16 rods were removed from similar HBU assemblies. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) Irradiated Fuels Examination Laboratory (IFEL) in early 2016 for detailed nondestructive examination (NDE) and destructive examination (DE). The detailed examinations will provide essential information on the physical state of the HBU rods and the fuel contained in the rods prior to loading, drying, and long-term dry storage. Similar tests will be performed at the end of the long-term storage period to identify any changes in the fuel rods’ properties during the dry storage period.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Instrumentation and Data Analysis Supporting the Transient Rod Pneumatic System Design Study at the Annular Core Research Reactor

The Sandia National Laboratories Annular Core Research Reactor (ACRR) has been experiencing intermittent failures of its pneumatically operated transient rod system. The most frequent failures have been to the dashpot rod where it connects to the upper aluminum connecting rod. Within the three systems, Transient Rod (TR) A has the most frequent failures. In order to conduct the pulse operations, the Transient Rods are pneumatically driven to quickly remove "poison" from the reactor core. During these pulse operations increased forces and vibrations are placed on the TR system. In an effort to fix these problems, Sandia's TA-V Nuclear Reactor and Systems Engineering Department (Engineering Department) has started an extensive Transient Rod Pneumatic System Design Study. The goal of this study is to determine the causes of the elevated failure rates in TR A. This necessitates an increased understanding of the dynamics in the TR systems. This report will cover the approach taken and reasoning behind the choices of instrumentation, calibration, data collection and data analysis used to diagnose these failures. In order to determine the types and specifications for the sensors that will be installed, estimates of the TR system performance needed to be determined. Using known and assumed parameters, mechanical and thermodynamic models were developed, and calculations performed to estimate ranges of displacement, speed, and acceleration experienced by the transient rod system. After some preliminary data collection, the thermodynamic models were refined for better accuracy. After installation of the full suite of sensors, data can be collected and analyzed in order to determine the cause of the increased failures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Sister Rod Destructive Examinations (FY2021) Appendix E Mechanical Testing

As a part of the 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 Zirc-4, ZIRLO, and M5. This report documents the status of the ORNL Phase 1 DE activities related to the mechanical testing of selected sister rods in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sister Rod Destructive Examinations (FY2021)

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. This report documents the status of the ORNL Phase 1 DEs of 7 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sister Rod Destructive Examinations (FY2021) Appendix F: Cyclic Integrated Reversible-Bending Fatigue Tests

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. This appendix documents the status of the ORNL Phase 1 DE activities related to tests of the sister rods using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Benchmark Exercise for the Control Rod Swelling Evaluation

The VTR core has six reactivity control assemblies and three safety assemblies. The control assemblies or primary control rods are adjusted during the normal operation to balance the core reactivity and to control the reactor power. A typical control assembly radial layout is presented in Figure 1. The figure shows the swelled absorber (B 4 C) rod. Initially, helium gas fills the gap between the pin and the cladding before irradiation swelling takes place. For VTR, HT9 steel was selected as the cladding and duct material. The main neutron absorbing material used in the VTR is B 4 C. When residing in the core, the neutronics, thermophysical, and mechanical properties of the materials used in a control assembly will degrade due to accumulated neutron damage. Material degradation limits how long a control assembly can reside in the core. Many phenomena affect the control assembly lifetime, such as the loss of reactivity worth due to B 4 C depletion, the mechanical interaction of the absorber rod and the cladding due to B 4 C swelling, the helium gas buildup in the pin due to B-10 capture, etc. B 4 C swelling, which causes closure of the gap between the absorber rod and the cladding, is usually considered as the main limiting factor from past experience. An initial study was conducted at PNNL to evaluate the irradiation behavior of a VTR control assembly. The evaluation was performed using the CNRD2 code that was initially developed for the FFTF. The study also included an assessment of the VTR control assembly and focused on a 61-pin control assembly design, which is different from that used (37-pin design) in the core design study. The study conducted by PNNL was reviewed independently by ANL. A Python script referred to as the Control Assembly Evaluation Script (CAES) was developed for the independent review and additional assessment of 37-pin control assembly design. The script has focused on the assessment of the absorber rod swelling for its importance in determining the control assembly lifetime. CAES uses geometry, neutronics, materials data as input to predict the swelling of the absorber rod during its residence in the reactor core. The results from CAES showed some non-negligible differences against the PNNL results. Some of the differences can be attributed to the different interpretation of the control rod assembly dimensions. To resolve this issue, a benchmark exercise was proposed. The benchmark specification was developed by PNNL. The benchmark exercise was performed independently at PNNL and ANL using different codes/scripts (CRND2 and CAES). This memo documents the results calculated using the different codes. However, this report is limited to presenting the results obtained. Further investigation of the cause of the observed difference will be performed as part of future activities, pending continuation of the VTR program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY22)

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. This report documents the status of the ORNL Phase 1 DEs of seven sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

High-Burnup Spent Fuel Data Project: Sister Rod Final Phase II Test Plan

The Sibling Pin test campaign is a Department of Energy (DOE) research activity within the Spent Fuel and Waste Science and Technology (SFWST) program that is tasked with characterization of high burnup (HBU) fuel in support of the High Burnup Spent Fuel Data Project. Of the 25 fuel rods in the Sibling Pin inventory, approximately 9 rod lengths have been consumed during the first phase (Phase I) of the test campaign leaving approximately 16 rod lengths for the second phase (Phase II) of testing. This plan outlines the Phase II testing and the motivations for performing these tests. Priorities for Phase II testing are based on previously identified knowledge gaps, lessons-learned from Phase I work, the original objectives of the High Burnup Spent Fuel Data Project and the Sibling Pin test campaign, and input from external stakeholders. The priorities for Phase II testing are to obtain data to characterize the effects of annealing on cladding mechanical properties and fuel rod performance, to quantify the creep behavior of cladding materials and fuel rods and the effects of creep deformations on the performance of cladding and fuel rods, and to gather data to support the final closure of the hydride reorientation and radial hydride induced embrittlement gap for HBU fuel rods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY23) Appendix F3: Uncertainty and Conservative Bias in the Cyclic Integrated Reversible-Bending Fatigue Test

The development and application of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) has been documented over the past 10 years. The CIRFT system was developed in collaboration with the U.S. Nuclear Regulatory Commission (NRC) to address the concern that high-burnup fuel rod fatigue performance was degraded compared to cladding-only behavior. Fatigue data measured by CIRFT on high-burnup fuel rods were then used by the NRC to develop a fatigue response curve for pressurized water reactor fuel rods with Zircaloy-4 cladding in NUREG-2224. Subsequently, CIRFT has been used in the sister rod project to test whether reoriented hydrides associated with vacuum drying influenced the fatigue lifetime, to extend the fatigue testing database to include fuel rods with ZIRLO cladding, and to add to the experience on both Zircaloy-4 and M5 cladding. These data will be used as a reference for subsequent characterization of fuel rods removed from the high-burnup demonstration cask to better understand the effects of long-term dry storage, which can be thought of as a multiyear elevated temperature anneal. The objective of fatigue testing is to provide data that can be used to develop a best-estimate fatigue curve and a conservative fatigue design curve that accounts for uncertainty and can be compared with actual transportation conditions. The process for defining a design fatigue curve for commercial used fuel rods was developed by Oak Ridge National Laboratory (ORNL) in FY22. One issue noted in the fatigue evaluation was that the calculated combined uncertainty in the strain amplitude values determined from the measured data was large. It was proposed then to directly measure the uncertainty in the strain amplitude to either validate the calculated combined uncertainty or update the uncertainty based on the measurement data. Thus, the purpose of this appendix is to document the measured uncertainty in the strain amplitude data. In the process, it was determined that there has been a conservative bias in the calculated strain amplitude that is small enough to be within the measurement uncertainty. This observation is discussed herein, and a case is made to continue to include the conservative bias in the data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Sister Rod Destructive Examinations (FY23) Appendix B: Segmentation, Defueling, Metallographic Data and Total Cladding Hydrogen

As a part of the 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. Both 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 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 of the SNF. This report documents the status of the ORNL Phase 1 DE activities related to: Rough segmentation (RS), Defueling (DEF), DE.02 optical microscopy (MET), and DE.03, cladding total hydrogen measurements. It is a cumulative update to the FY22 status report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗