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DOE OSTI2016

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Narrowing transient testing pulse widths to enhance LWR RIA experiment design in the TREAT facility

The Transient Reactor Test (TREAT) Facility provided thousands of transient irradiations and plays a crucial role in nuclear-heated safety research. The facility’s flexible design and multi-mission nature saw historic experiments for numerous reactor fuels and transient types but was never specifically adapted to address very-brief pulse transients akin to postulated Light Water Reactor (LWR) Reactivity Initiated Accidents (RIA). Since the behaviors of fuel under these conditions depends strongly on energy input duration and resulting cladding time/temperature response under pellet cladding mechanical interactions, this three-year project was conceived to investigate new pulse-narrowing capabilities. Kinetic models showed incremental improvements for minor facility enhancements including increased reactivity step insertions (to initiate the power pulse) and slightly-increased transient rod drive speed for pulse termination (“clipping”). Replacing peak fuel assemblies, repositioning non-transient control rods to hold down the “hot side” of the core, and balancing against required excess reactivity needs, the limiting fuel assembly power can be reduced by ~20 %. Furthermore, this is a remarkable discovery of latent capability in TREAT, not only in enabling the subject capabilities for reduced pulse widths, but also significantly “uprating” the core’s transient energy capacity. Incremental improvements can likely enhance TREAT’s capability into BWR-relevant missions; briefer PWR pulse shapes can only be achieved with an advanced clipping system. Helium-3 ( 3 He) was found to offer the greatest benefits for clipping design and overall performance. A unique concept showed great promise for enabling a 3 He-based system with a reasonable cost; this design concept will likely become the focal point of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Multiphysics analysis of fuel fragmentation, relocation, and dispersal susceptibility–Part 1: Overview and code coupling strategies

The US nuclear energy industry is investigating strategies to increase the reactor operating cycle to 24 months, resulting in peak rod average burnups exceeding the current limit of 62 GWd/tU. This increase will in turn increase the probability of fuel fragmentation, relocation, and dispersal (FFRD) in the event of a loss-of-coolant accident (LOCA). This effort couples multiple codes to (1) evaluate full-core power histories for high-burnup fuel operated in a Westinghouse 4-loop pressurized water reactor, (2) model a postulated large-break LOCA, and (3) calculate the mass of fuel susceptible to FFRD. This paper, the first of three describing the work, focuses on code coupling strategies and FFRD susceptibility calculations. The other two companion papers focus on code-specific designs and analyses. Three codes were used in this work. VERA was used to calculate steady-state power histories, TRACE was used to model the transient thermal hydraulics, and BISON was used to model steady-state and transient fuel performance and cladding failure. Herein, several fuel pulverization models were used to calculate FFRD susceptibility in failed rods. Depending on the cladding failure/fuel pulverization model combination, the core-wide FFRD susceptibility during the postulated LOCA range from 0 to over 5,000 kg.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. Further, a large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Mechanical degradation of duplex SiC-fiber reinforced SiC matrix composite tubes under a controlled high-temperature steam environment

This paper describes results from the mechanical evaluation of unirradiated SiC fiber–reinforced SiC matrix composite tubes under a controlled high-temperature steam environment. The experiments were aimed at identifying key material degradation behavior under environments relevant to loss-of-coolant accidents of light water reactors. Mechanical tests of the SiC composite tubes at 1000 °C under steam and inert environments were conducted using a unique test capability. Further, the material tested was a duplex tube with a thick monolithic SiC layer on the outer surface. The tubes were subjected to preloading at ~100 MPa in tension before exposure to high-temperature steam with up to 75% of the preload at a constant displacement. In the presence of matrix cracks, the steam exposure caused embrittlement of the SiC composite tubes and failure at a stress level below the pretest stress. The material degradation was explained by a fiber oxidation model, which can be applied to various SiC cladding concepts. The embrittlement could be a limiting factor for using SiC cladding subjected to loss-of-coolant accident conditions.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Qualitative and quantitative analysis of neutron irradiation effects in SiC/SiC composites using X-ray computed tomography

Silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composites are candidate materials for cladding of light water reactor (LWR) fuels. Loss of fission product gas retention due to the formation of microcrack networks is considered a potential failure mechanism for SiC/SiC-cladded fuels. In this work, a variety of SiC/SiC composite tubes were irradiated with and without an LWR-relevant radial heat flux in the High Flux Isotope Reactor, followed by detailed characterization with X-ray computed tomography (XCT). This first set of XCT data for neutron-irradiated samples confirmed that the internal stresses arising from a combination of temperature gradients and irradiation-induced swelling act as the primary driver for cracking. Consequently, while the observed cracking patterns varied depending on the tube architectures, the sharp edges of relatively large pores were found to be the common stress concentrator. These findings are useful to help improve the design and manufacturing of SiC/SiC fuel claddings for reduced failure probability.

42 ENGINEERING↗

DOE OSTI2020

Modelling uranium dioxide corrosion under repository conditions: A pore-scale study of the chemical and thermal processes

This paper investigates the corrosion behaviors of spent uranium dioxide (UO 2 ) fuel when placed in geological repositories. We performed pore-scale reactive transport simulations of the UO 2 corrosion process in a defective fuel rod with different orientations of breach or fracture on its clad. It is found that the corrosion rate has strong dependency on the breach orientation. The highest corrosion rate of UO 2 is calculated when the angle between the groundwater flow direction and the clad fracture reaches 180°. UO 2 corrosion simulations with different flow rates are compared. The results show the higher flow rate accelerates the corrosion of UO 2 fuel. The effect of pH on the corrosion process is also determined. It is indicated that higher pH can limit the corrosion of UO 2 by reducing the reaction rate. The dependence of reaction rate and time on reactive surface area is explored. Spent fuel with lower surface area demonstrates longer lifetime under corrosion conditions. A numerical model coupling thermal conduction and chemical reactions is developed to assess the impact of temperature on the process of UO 2 corrosion. The results show that higher temperature leads to larger corrosion rates for UO 2 . Here, the predicted reaction rates are higher in comparison with the isothermal results. Hence, our results can help improve the fundamental understanding of UO 2 corrosion in geological repositories for long-term storage of spent nuclear fuels and provide guidance for the safe operations and selection of appropriate repositories.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Analysis of inconel 600 oxidized under loss-of-coolant accident conditions: A multi-modal approach

Understanding the performance of cladding materials during a loss-of-coolant accident scenario is crucial for developing next-generation accident tolerant nuclear fuel rod claddings. We describe a multi-modal approach combining synchrotron-based diffraction, and Raman spectroscopy with nano-scale electron microscopy techniques for investigating oxidation of Inconel 600 (A600) in steam and air environments at 1200 °C for 2 h. We report that A600 exposed to steam develops a Cr 2 O 3 -enriched surface layer while a mixed phase oxide layer containing NiFe 2 O 4 , Cr 2 O 3 , Fe 3 O 4 and Fe (3–x) CrxO 4 is formed in air. Mechanism of oxidation for A600 in air and steam environments is discussed.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Determining failure properties of as-received and hydrided unirradiated Zircaloy-4 from ring compression tests

An integrated experimental and computational approach was developed to determine the plastic strain at fracture and to investigate the fracture energy of as-received and hydrided unirradiated Zircaloy-4 nuclear cladding tubes in the hoop direction at room temperature from ring compression testing (RCT). This work builds on previous methods in the literature that were developed to obtain the mechanical properties of nuclear cladding before fracture (Young’s modulus, yield stress, and strain-hardening parameters) from experimental RCT results and extends the characterization of this material class to include material failure properties. Furthermore, a ductile damage approach was developed and implemented in a validated finite element model to predict material failure, evaluate fracture energy, and quantify the plastic strain at fracture initiation. The hydrogen content of the Zircaloy-4 specimens was varied up to 610 wppm when the failure became brittle. As expected, increasing hydrogen content caused embrittlement of the Zircaloy-4 samples tested and the plastic strain at fracture and fracture energy to decrease.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M&S capabilities, which is in progress, will assist in addressing long-standing and future operational and safety challenges of the nuclear industry. - Highlights: • Complexity of physics based modeling of light water reactor cores being addressed. • Capability developed to help address problems that have challenged the nuclear power industry. • Simulation capabilities that take advantage of high performance computing developed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Multimodal analysis and characterization of the boehmite layer formed on AA6061 before and after alkaline etching

Low-enriched uranium (LEU) alloyed with 10% Mo (U-10Mo) is being considered as a promising alternative to oxide-based dispersion fuel with high-enriched uranium for use in research reactors. The configuration of this proposed LEU monolithic LEU fuel plate consists of a U-10Mo plate-type fuel foil with a 25 µm zirconium interlayer barrier clad with an aluminum alloy (AA6061). In certain research reactors, the clad AA6061 is coated with a boehmite layer to prevent corrosion. The boehmite layer has a high-pH passivation range, which makes it resistant to oxidation. Boehmite is usually formed on the AA6061 surface by autoclave processing. Before the boehmite layer is added, the surface of the AA6061 is cleaned using techniques such as polishing and wet etching. In this study, we use multimodal analysis to examine how pretreatment of AA6061 using polishing followed by alkaline etching affects the chemical composition of the boehmite layer. X-ray photoelectron microscopy (XPS), transmission electron microscopy (TEM), and x-ray diffraction (XRD) were used to study the chemical changes in the boehmite layer caused by alkaline etching pretreatment. XPS provides quantitative analysis for the Al:O ratio as well as oxidation states present on the surface, which suggests slight oxidation of the boehmite surface after alkaline etching of the AA6061 surface. We further explored this suggested oxidation of the boehmite surface using high-resolution transmission electron microscopy with selected area electron diffraction (SAED) and grazing incidence x-ray diffraction (GI-XRD), which suggested only a small amount of aluminum oxide at the surface. The multimodal analysis and imaging yielded new insights for optimizing boehmite growth on AA6061 for research reactors.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Uniform corrosion of FeCrAl alloys in LWR coolant environments

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. Finally, the maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ~2 μm, which is inconsequential for a ~300–500 μm thick cladding.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Microstructure studies of interdiffusion behavior of U 3 Si 2 /Zircaloy-4 at 800 and 1000 °C

Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U 3 Si 2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000°C up to 100 hours. The microstructure of pristine U 3 Si 2 and U 3 Si 2 /Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800°C is ZrSi 2 , with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi 2 interface and Fe-Cr-U-Si phases at ZrSi 2 /U-Si interface. As a result, the primary interdiffusion products at 1000°C were Zr 2 Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U 6 Fe.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Solid-liquid phase equilibria of Fe-Cr-Al alloys and spinels

Ferritic FeCrAl alloys are candidate accident tolerant cladding materials. There is a paucity of data concerning the melting behavior for FeCrAl and its oxides. Analysis tools have therefore had to utilize assumptions for simulations using FeCrAl cladding. The focus of this study is to examine in some detail the solid-liquid phase equilibria of FeCrAl alloys and spinels with the aim of improving the accuracy of severe accident scenario computational studies.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2 O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Young's modulus evaluation of high burnup structure in UO 2 with nanometer resolution

The mechanical properties of a well-characterized irradiated light water reactor fuel pin with an average burnup of 72 MWd/kgU were investigated by utilizing nano-indentation technique. Young's modulus and hardness are reported as functions of radial positions covering from mid radial position in the urania pellet to high burnup structure, fuel-cladding chemical interaction zone and zirconium-based cladding. Here, by probing microstructure of the high burnup zone, the mechanical properties of sub-grains and the interaction between them were examined at the nanometer scale. Force-displacement curves from each of these individual sub-grains suggest they resemble single crystal UO 2 . However, these restructured grains are weakly bonded to their neighbors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Characterization of surface layers formed on DU10Mo ingots after processing steps and high humidity exposure

The design of monolithic U-Mo fuel elements fabricated from low-enriched uranium for use in high-power research reactors requires bonding of the fuel foil to either Al cladding or a Zr barrier layer. Processing of the U-Mo ingot to final foil form has the potential to generate surface layers on the foil that differ from the bulk, metallic U-Mo. The interfacial properties between the U-Mo and Zr or Al cladding layers will then be determined by these surface layers. We use x-ray photoelectron spectroscopy, cross-sectional scanning electron microscopy, and atomic force microscopy to characterize the composition, oxidation state, and morphology of the surface layers that form after hot rolling and cold rolling depleted U–10 wt% Mo alloy (DU10Mo). A thick uranium nitride layer is observed after hot rolling, although its origin is likely from a previous processing step. The surface composition of cold-rolled DU10Mo varied significantly between two foils, with a strongly Mo-depleted surface oxide layer observed on one foil but nominal Mo composition observed on the other. The efficacy of acid etching in HNO 3 is compared to that of electropolishing in H 2 SO 4 to remove the surface layers, and both methods are found to be similarly effective. Both laboratory (low humidity) air exposure and longer rinse times in water are shown to promote the formation of surface oxide layers. Exposure of both acid-etched and electropolished DU10Mo foils to humid air (97% relative humidity) for six weeks results in formation of a thick oxide layer due to corrosion. The oxide layer on the acid-etched foil is thicker and more highly oxidized than the oxide layer that forms on the electropolished foil, and these differences in oxidation behavior are attributed to higher surface roughness on the acid-etched foil. In general, Mo is found to play a role as a sacrificial element, typically exhibiting a larger ratio of Mo 6+ / Mo 4+ than U 6+ / U 4+ . This is unexpected, given the greater thermodynamic driving force to form U oxides than Mo oxides.

Kaspar, Tiffany C. (ORCID:0000000328167569)↗

DOE OSTI2018

Evaluation of post-weld heat treatments applied to FeCrAl alloy weldments

The nuclear incident at the Fukushima Daiichi nuclear power plant has created a strong push for accident-tolerant fuel cladding to replace current zirconium-based cladding. A current near-term focus on iron-chromium-aluminum (FeCrAl) alloys. Laser-welded FeCrAl samples (C35MN, C37M, and C35M10 TC) were subjected to three different post-weld heat treatment regimes: 650 °C for 5 h, 850 °C for 1 h, and 850 °C for 5 h. Here, the samples were then analyzed using optical light microscopy, micro-hardness indentation, and scanning electron microscopy coupled with energy-dispersive spectroscopy and electron backscatter diffraction. The base microstructure of C37M and C35M10 TC experienced significant grain coarsening outside the fusion zone due to the applied post-weld heat treatments, whereas Nb-rich precipitation in C35MN limited grain growth compared with the other alloys studied.

36 MATERIALS SCIENCE↗