NEAMS / AFC High Burnup Coordination Project - Assessment
This short presentation summarizes the NEAMS/AFC collaboration on assessment activities for high burnup applications.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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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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Summary of physics-based pulverization model development for high-burnup fuel
Summary to be presented at EPRI-DOE workshop on high burnup and accident tolerant fuel
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High-burnup Experiments in Reactivity Initiated Accidents (HERA)
HERA is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under reactivity-initiated accidents (RIA). During RIAs the fuel cladding can be breached by one of two mechanisms; pellet-cladding mechanical interaction or high temperature balloon and rupture. Pellet-cladding mechanical interaction (PCMI) failure occurs in the early phase of an RIA transient where the fuel pellet expands rapidly into the cladding prior to any meaningful heat transfer from the pellet to the cladding. This results in high stresses in the cladding resulting in crack propagation through the base Zircaloy metal and axially along the cladding tube. Failure by PCMI is a brittle failure mode and occurs at low cladding strains, approximately 2% hoop strain. If the cladding survives the PCMI phase of the transient, it may still fail due to ballooning and rupture induced by extended exposure to high temperature. Thermal transport from the fuel pin to the coolant is significantly reduced if a transition to film boiling on the cladding surface occurs. This result in both a rapid increase in fuel temperature, which drives fission gas release, and a spike in cladding temperature, which results in loss of mechanical strength. If the internal pressure of the fuel rod is greater than the system pressure the cladding can inelastically deform and eventually rupture. Failure by swelling and rupture occurs at larger strains, generally greater than 5% hoop strain. While PCMI failures are generally seen as more limiting, the dominance of one failure mode over the other can be affected by both the cladding material conditions, and the transient evolution.
(HERA)↗
HIGH BURNUP FUEL-COOLANT INTERACTION ANALYSIS SUPPORTING FUEL SAFETY TESTING AT IDAHO NATIONAL LABORATORY
In the near future, experiments on HBu fuel under loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) conditions will be performed within the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL). These experiments will be performed using the Transient Water Irradiation System for TREAT (TWIST) experiment vehicle. To support these experiments, analysis of fuel-coolant interaction (FCI) energetics is underway. This paper discusses FCIs in the context of light water reactor (LWR) safety, differentiating between the severe accident focus of commercial reactors and experimental RIA test programs where FCIs have occurred. However, it is highlighted that as the nuclear industry aims for increased burnup limits, the FCI events observed in RIA test programs may become relevant to commercial LWR safety analysis. The paper then presents developments to the UW-FCI computer program to enable simulation of FCIs initiated by solid fuel particles dispersing into the coolant during RIAs.
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(S)TEM/EDS study of native precipitates and irradiation induced Nb-rich platelets in high-burnup M5®
We have investigated microstructure and microchemistry of precipitates and dislocation loops in high-burnup M5® using (scanning) transmission electron microscopy ((S)TEM) equipped with energy dispersive X-ray spectroscopy (EDS). Two (S)TEM lamellae were made by cryo-FIB from the same cladding sample. The Nb-rich native precipitates were found in the metal, in the suboxide and in the oxide. Upon diffraction analysis, most of the Nb-rich native precipitates in the metal matrix remain as β-Nb phase, while no β-Nb native precipitates were found in the oxide. Nearby the oxide and metal (O/M) interface, the native precipitates in the oxide were already oxidized into t-NbO2 phase. At further distance away from the O/M interface, the Nb-rich native precipitates were gradually oxidized and became amorphous. Besides the native precipitates, Nb-rich irradiation induced precipitates (IIPs) were found in the metal matrix. Using g = <0002> vector for imaging, the length of the IIPs was aligned with dislocation loops. However, no Nb segregation to the dislocation loops themselves was observed. For the first time, we report that IIPs indeed exist in the oxide but only within about 1.5 µm away from O/M interface. However, the oxidation state of the IIPs in the oxide is still unclear. The presence of both native precipitates and IIPs in the oxide may indicate the Nb concentration in the oxide solid solution remain low nearby the O/M interface, which may explain the reduced corrosion kinetics of in-pile M5®. On the other hand, no IIPs were observed in the oxide at further distance and this may indicate that they have eventually dissolved back into the oxide. A few mechanisms related to IIPs stability in the oxide are presented.
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Electron microscopy characterization of the fuel-cladding interaction in medium burnup annular fast reactor MOX
In this work, we present electron microscopy data focused on the fuel-cladding interaction layer in annular fast reactor MOX with HT-9 cladding at medium burnup. In agreement with previous literature data, the volatile fission products Cs, Te and I have migrated radially into the extreme fuel periphery and partially interacted with the cladding. The accumulation of Cs has occurred in the outermost rim of the fuel pellet, where grain recrystallization has also been observed. Significant amounts of Pd have been found in the interaction zone, particularly in the sample taken from the upper half of the fissile column where the cladding temperatures are higher. At this axial location, Cr has been enriched at the cladding inner surface and diffused into the fuel. Furthermore, chromium remains mainly in metallic form, but locally formed oxides. The fission products Cs, Te and I are found, with variable composition, in form of nanocrystalline regions dispersed in the metallic Cr-rich layer. The morphology and chemical characteristics of the layer suggest a non-oxidative corrosion mechanism as principal cladding degradation phenomenon occurring in this sample, with local onset of Cr oxidation within the nanocrystalline precipitates.
36 MATERIALS SCIENCE↗
Post-irradiation examination of low burnup U 3 Si 5 and UN-U 3 Si 5 composite fuels
Here this work presents post-irradiation examination data on UN-U 3 Si 5 and U 3 Si 5 fuels at low burnup (i.e., <10–15 GWd/tHM) with Kanthal AF® cladding. The results suggest good irradiation performance for both the silicide and nitride-silicide composite pellets. Optical microscopy revealed that the pellet-cladding gap is still open, and limited axial cracking was observed only in UN-U 3 Si 5 pellets. Microcracking was isolated to the U 3 Si 5 phase in all cases and was observed in pre-irradiation and depleted pellets, indicating that it was not irradiation induced. The fission gas release was minimal for the calculated fission density achieved (2.6 – 3.15 × 10 20 fiss/cm 3 ). No fission gas bubbles were observed in the optical metallography. These results suggest acceptable swelling and fission gas behavior for both the single phase and composite compositions.
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Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance
Here, the development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.
BISON↗
Elucidating the effect of minor-actinide addition on fuel-cladding chemical interaction in an HT-9 clad U-Pu-Zr metallic fuel irradiated to 6.15 at.% burnup in EBR-II
Scanning and transmission electron microscopy (S/TEM) were used to characterize the local fuel-cladding chemical interaction (FCCI) in one cross-section taken from a HT-9 clad U-20.3Pu-10Zr-1.2Am-1.3Np (in wt.%) fuel irradiated to 6.15 at.% burnup with inner cladding temperatures ranging between 460–490 °C. Results showed that the total interaction thickness between fuel and cladding was <10 µm. Fe infiltrated the fuel to form U-Zr-Fe phases while fuel elements or lanthanides did not infiltrate into the cladding. Np was not involved in the formation of any phases in the examined locations; however, Am played a role by forming a ∼2 µm thick homogeneous Fe-Pu-Am planar front at the inner cladding wall. An oxidized Na layer existed in the fuel-cladding gap with Fe and lanthanide particles dispersed within, suggesting Na could facilitate the transport of fuel and cladding constituents. Secondary phases, including an FCC Zr-rich phase, lanthanide phases, and α’-Cr(Fe) were identified in the outer fuel and FCCI regions. Furthermore, this study suggests that, for the irradiation conditions specific to this cross-section, minor actinides have little impact on FCCI behavior beyond what would be observed in typical HT-9 clad U-Pu-Zr fuel pins systems.
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Using X-ray spectrum imaging to quantify fission gas in high-burnup UO2 fuel
Analysis of fission gas bubbles (FGBs) in light-water reactor (LWR) fuel is needed to improve the understanding and predictive capability of fuel evolution under normal- and transient conditions. One of the most important parameters of a FGB is the pressure of the gas, primarily Xe, inside it. Bubble volume and location (inter- vs. intragranular) are important considerations as well. However, experimental analysis of such bubbles is challenging due to their small size and embedded nature, and usefulness of the data requires large numbers be analyzed. This paper proposes a method to measure the pressure of Xe bubbles using X-ray spectrum imaging (XSI) in scanning transmission electron microscopy (STEM). From X-ray generation and instrumental parameters, the number of Xe atoms yielding a given number of Xe L-series X-ray counts is estimated, and then from the estimated volume of the bubble the Xe density is estimated, which can then be converted to pressure via an equation-of-state. We apply the method to XSI datasets from high burnup (HBu) fuel from North Anna 1 reactor in as-irradiated and post-LOCA-test conditions and find Xe pressures in the FGBs clustered around 1 GPa.
Parish, Chad [ORNL] (ORCID:0000000312097439)↗
Characterization of Fuel Cladding Chemical Interaction on a High Burnup U-10Zr Metallic Fuel via Electron Energy Loss Spectroscopy Enhanced by Machine Learning
Fuel cladding chemical interaction (FCCI) is one of the main performance limiting factors for metallic nuclear fuels. The interaction destabilizes the martensitic microstructure and deteriorates mechanical properties of HT-9 cladding. The detection of low atomic number elements (Z<10) and overlapping of elemental peaks can be problematic in interpreting energy dispersive X-ray spectroscopy (EDS) data. Electron energy loss spectroscopy (EELS) provides precise elemental edge energy values and can detect elements with a low atomic number. This work utilizes EELS to study the distribution of lanthanides and light elements at the interaction region. The sample was prepared from the FCCI region of a U-10Zr (wt.%) solid fuel with HT-9 cladding, irradiated to a burnup of 13.2 at.%. Processing the EELS data included three major steps: 1) enhance the signal to noise ratio by denoising the spectrum with principal component analysis (PCA) method, removing background and performing deconvolution; 2) identify chemical elements with core energy loss edges; 3) confirm different phases using a popular machine learning method, K-means. This work presents qualitative assessment of lanthanides and light elements like carbon (C) and oxygen (O) enhanced by the application of machine learning algorithms. By comparing with EDS elemental maps, EELS provides higher resolution chemical maps, reveals the distribution of carbon at the interaction region supporting the formation of zirconium carbide, a rind-like microstructure feature that was proposed to mitigate the chemical interaction. Furthermore, the plasmon peak map was also found to indicate an energy shift associated with the formation of phases/compounds. K-means clustering method was used on the processed electron energy loss (EEL) spectrum to automatically reveal different phases. The resulting clustered maps from K-means clustering align well with elemental maps confirming certain phases, especially Fe-Ce and Zr-C, in the FCCI region.
EELS↗
Effects of grain size and porosity on cladding failure in high-burnup UO 2 : A sensitivity and uncertainty study
Isotopic taggants are being studied to aid in the provenance assessment of nuclear materials. However, these taggants must be selected such that they do not adversely affect fuel performance during normal operation or accident scenarios. Taggants are known to affect the fuel’s grain size and porosity. In the work described in this paper, the BISON fuel performance code was used to assess the potential effects of taggants (i.e., grain size and porosity) on fuel rod behavior and cladding failure during a high-burnup, large-break loss-of-coolant accident. Here, 281 individual fuel rods from the same reactor core were modeled for a sensitivity study, a parametric study, and uncertainty quantification. The cladding failure predictions often exhibited stochastic behavior. After additional study, it was found that the cladding failure model is highly sensitive to residual error inherent to numerical approximation solvers. Some strategies to mitigate this sensitivity are discussed. The study found no relationship between known taggant effects and cladding failure status. However, taggants were found to affect the time and location of failure in certain rods. In conclusion, future work to continue investigating and validating these findings is briefly discussed.
Doped UO 2↗
The need for speed – Burnup determination of spent nuclear fuel
Here we present the application of a novel analytical approach for high-speed evaluation of burn-up in spent nuclear fuel. This automated online separation–direct analysis method, which simultaneously measures the isotopic compositions and concentrations of Nd, Pu, and U using isotope dilution-inductively coupled plasma mass spectrometry, enables the determination of burnup in as little as 4.5 h per sample with expanded uncertainties as low as 1.2% using as little as 1µg of fuel. We compare the determined burn-up values, sample requirements, and analysis times with an ultra-high precision, in-house developed, methodology which employs multiple analytical techniques to measure the isotopic compositions and concentrations of Nd, Pu, and U yielding burn-up values with expanded uncertainties as low as 0.49% at the 95% confidence level, the lowest reported to date.
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Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask
This report provides a best-estimate evaluation of residual water content (post-dry out) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask, and evaluates the radiolysis of the residual free water, and the physisorbed and chemisorbed waters on the surfaces of the fuel and cask internal contents. The evaluation of radiolytic breakdown of those waters with gamma radiation causing the generation of hydrogen gas (H 2 ) is made using available literature data and models. This evaluation is part of the overall materials performance evaluation of the SNF-in-canister system, and is part of the technical bases for their continued safe dry storage. The TN-32 cask contents included 32 HBU LWR spent fuel assemblies each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless steel structural components. The residual free and surface (physisorbed/chemisorbed) waters are ascribed to water vapor in the free volume and to components’ surfaces, respectively. The total potential radiolytic hydrogen inventory from the water vapor and from waters ascribed to surfaces has been calculated assuming all the water produced molecular H 2 . The residual water that is chemically incorporated into the bulk of a hydrated oxide, i.e., chemisorbed water, and its total potential hydrogen inventory has been calculated. These calculations are at the physical limit of material available and are used for a bounding assessment purpose only.
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Sister Rod Destructive Examinations (FY20) Appendix D: Fission Gas, Fuel Burnup and Fuel Isotopic Measurements
This appendix documents the status of the ORNL Phase 1 DE activities related to fission gas sample analysis, fuel burnup analysis, and fuel isotopic analysis of selected sister specimens in Phase 1 of the sister rod test program.
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Engineering Assessment of UO 2 and Cladding Behavior under High Burnup LOCA Conditions
To maximize the data extracted from a limited number of high-burnup fuel rod samples, several modeling efforts were performed to elucidate the fuel and the cladding responses of these fuels under transient conditions. These objectives were (1) to determine the role of the fuel stress state in fuel pulverization, (2) to ascertain the differences between conditions of cladding burst during experiments and those expected during a commercial reactor transient, and (3) to develop a method to conservatively calculate the geometry of the cladding rupture’s opening to inform fuel dispersal susceptibility.
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Assessment of Modeling and Simulation Technical Gaps in Safety Analysis of High Burnup Accident-Tolerant Fuels
The United States nuclear industry is facing a strong challenge to maintain regulatory-required levels of safety while ensuring economic competitiveness to stay in business. Safety remains a key parameter for all aspects related to the operation of light water reactor (LWR) nuclear power plants (NPPs), and it can be achieved more economically by using a risk-informed ecosystem, such as that being developed by the Risk-Informed Systems Analysis (RISA) Pathway under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. The LWRS Program is promoting a wide range of research and development activities to maximize both the safety and economically efficient performance of NPPs through improved scientific understanding, especially given that many plants are considering second license renewal. The RISA Pathway has two main goals: The deployment of methodologies and technologies that enable better representation of safety margins and the factors that contribute to cost and safety, and; The development of advanced applications that enable cost-effective plant operation. As part of the RISA Pathway, the Enhanced Resilient Plant (ERP) project refers to an NPP where safety is improved by implementing various measures, such as accident-tolerant fuels (ATF), diverse and flexible coping strategy (FLEX), enhancements to plant components and systems, incorporation of augmented or new passive cooling systems, and utilization of advanced battery technologies. The objective of the ERP research is to use novel methods and computational tools to enhance existing reactors’ safety while reducing operational costs. Many U.S. utilities are targeting implementation of ATFs instead of traditional fuel in the near future since ATFs offer benefits in terms of improved performance and cost savings. The robust properties of ATF make it possible to extend the refueling cycle from 18 to 24 months in addition to the opportunity to use less of fuel. Extensive safety assessments are required to support regulatory requirements and obtain the approvals to use ATFs and the ERP project support the industry by developing novel effective methodologies for safety evaluations. In this project, the technical gaps in the modeling and simulation of the high burnup (HBU) ATF were assessed in terms of the fuel cladding behavior during the postulated accident events. The issues were identified in modeling the cladding deformation, the hydrodynamic change due to cladding deformation and the critical heat flux (CHF). The RELAP5-3D cladding deformation model was assessed by multiple verification tests and validation with the instrumented fuel assembly (IFA) experiment.
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Safety Analysis of FeCrAl Accident-Tolerant Fuels with Increased Enrichment and Extended Burnup
The U.S. nuclear industry is facing a strong challenge to maintain regulatory-required levels of safety while ensuring economic competitiveness to stay in business. Safety remains a key parameter for all aspects related to the operation of light water reactor nuclear power plants (NPPs), and it can be achieved more economically by using a risk-informed ecosystem, such as that being developed by the Risk-Informed Systems Analysis Pathway under the U.S. Department of Energy Light Water Reactor Sustainability Program. This program is promoting a wide range of research and development activities to maximize both the safety and economically efficient performance of NPPs through improved scientific understanding, especially given that many plants are considering a second license renewal. The Risk-Informed Systems Analysis Pathway has two main goals: (1) The deployment of methodologies and technologies that enable a better representation of the safety margins and factors that contribute to cost and safety, and (2) The development of advanced applications that enable cost-effective plant operation. As part of this pathway, the Enhanced Resilient Plant project refers to an NPP where safety is improved by implementing various measures, such as accident-tolerant fuels, diverse and flexible coping strategies, enhancements to plant components and systems, incorporation of augmented or new passive cooling systems, and utilization of advanced battery technologies. The objective of the Enhanced Resilient Plant project is to use novel methods and computational tools to enhance existing reactors’ safety while reducing operational costs. This report documents research and development conducted in support of deployment of accident-tolerant fuels. This project performed safety analyses for the steady-state normal operation, anticipated operational occurrences, and design-basis accidents of a representative four-loop pressurized water reactor model with Zr and FeCrAl accident-tolerant fuel clad with higher enrichment and burnup supporting plant refueling cycles of 18 and 24 months. The source terms and environmental impacts were studied for a large-break loss of coolant accident, including uncertainty analyses.
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