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DOE OSTI2014

Irradiation Creep in Graphite

An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

Ubic, Rick↗

DOE OSTI2015

Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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

Analysis of Cavern and Well Stability at the Bryan Mound SPR Site Using the M–D Salt Creep Model

This report presents computational analyses that simulate the structural response of caverns at the Strategic Petroleum Reserve Bryan Mound site. The cavern field comprises 20 caverns. Five caverns (1, 2, 4, and 5; 3 was later plugged and abandoned) were acquired from industry and have unusual shapes and a history dating back to 1946. The other 16 caverns (101-116) were leached according to SPR standards in the mid-1980s and have tall cylindrical shapes. The history of the caverns and their shapes are simulated in a 3-D geomechanics model of the site that predicts deformations, strains, and stresses. Historical wellhead pressures are used to calculate cavern pressures up through July 2016. Because of the extent of heterogeneous creep behavior observed throughout the Bryan Mound site, a set of cavern-specific creep coefficients was developed to produce better matches with measured cavern closure and surface subsidence. For this new implementation of the model, there are two significant advances: the use of the multimechanism deformation (M-D) salt creep model to evaluate both steady-state and transient salt creep; and the creation of finite element mesh geometries for the caverns that nearly exactly match the geometries obtained through sonar measurements. The results of the finite element model are interpreted to provide information on the current and future status of subsidence, well integrity, cavern stability, and drawdown availability.

58 GEOSCIENCES↗

DOE OSTI2018

The Initiation of Long-Term Creep Rupture Tests on the First Alloy 709 Commercial Heat

A systematic creep tests were conducted as part of a multi-lab effort to identify the material conditions for Alloy 709 Code Case Project. Nine candidate material conditions were studied with three melt conditions each with three S/A temperatures 1050, 1100 and 1150°C. The three melt conditions included AOD, ESR and ESR + Homogenization. The screening creep rupture tests were conducted under 600°C, 330MPa. AOD 1100 and ESR 1100 have been identified to be the top two material conditions for the Code Case project after a comprehensive evaluation. Long-term creep tests were initiated on AOD 1100 at the ORNL. The six creep tests were started at temperatures from 550 to 800°C, stresses from 38 to 309 MPa to target rupture time 60,000 hours. The loading strains agreed with those obtained from tensile tests at the corresponding temperatures.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Thermo-mechanical Analysis of SiC/SiC Composite Cladding with BISON including Fuel Creep

This report presents the results of a parametric thermo-mechanical parametric analysis of SiC/SiC cladding performed using the BISON fuel performance code. In this analysis different steady state linear heat rates (LHR) (18kW/m, 22kW/m, 26kW/m and 30kW/m) and initial fuel rod-gap thicknesses (50μm, 65μm, 80μm and 95μm) were considered along with three different model implementation of fuel creep. These include 1) no fuel creep 2) fuel creep based on the Hagrman model and 3) fuel creep based on a modified model. The fuel rod geometries used in this analysis are representative of a generic 17 x 17 fuel rod assembly used in a Westinghouse PWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Stress and tangent update equations for combined time-hardening creep and J 2 plasticity in an implicit hypo-elastic formulation

This work is motivated by the need to analyze the behavior of metallic nuclear fuels which under normal operating conditions build up stresses due to non-homogeneous thermal expansion, fission gas and solid product swelling among other phenomena that are simultaneously relaxed by creep and plastic flow. This report details the stress and tangent update equations for combined J 2 based rate independent plasticity and time-hardening creep effects in a fully implicit hypo-elastic formulation involving two cases: pure creep without plasticity where the yield criterion has not yet been met and the combined effect of both creep and plasticity beyond yield. Closed form expressions for the consistent material tangent to be used in both cases are derived which can be used in implicit codes and is expected to help in obtaining optimal convergence rates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

A general constitutive framework for the combined creep, plasticity and swelling behavior of nuclear fuels in an implicit hypoelastic formulation

This work is motivated by the need to analyze the behavior of nuclear fuels which under normal operating conditions build up stresses due to non-homogeneous thermal expansion, fission gas and solid product swelling among other phenomena that are simultaneously relaxed by creep and plastic flow. This report details the stress and tangent update equations for combined J2 based rate independent plasticity, timehardening creep and fission gas induced swelling e ects in a fully implicit hypo-elastic formulation involving two cases: pure creep without plasticity where the yield criterion has not yet been met and the combined e ect of both creep and plasticity beyond yield. Closed form expressions for the consistent material tangent to be used in both cases are derived which can be used in implicit codes and is expected to help in obtaining optimal convergence rates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

X-ray characterization of atomistic defects causing irradiation creep of SIC

Silicon carbide (SiC) fiber–reinforced SiC matrix composites have been widely studied for use in developing an accident-tolerant fuel (ATF) cladding for light water reactors (LWRs) in recent years because of their superior oxidation and irradiation resistance and low neutron absorption cross section. An important task is the establishment of the thermo-mechanical analysis capability to predict the in-pile performance and failure probability of the cladding. Irradiation creep is known to be a key property in modeling the stress state under irradiation. To better understand the radiation creep mechanism from an atomic point of view, this study conducted synchrotron-based X-ray diffraction (XRD) experiments on SiC neutron-irradiated with and without applied stress at the National Synchrotron Light Source-II (NSLS-II). This report presents progress of the XRD study and future plans. The important finding obtained was that the macroscopic creep strain and strain due to the lattice expansion were similar in case of neutron irradiation at 300°C to 0.1 displacement per atom (dpa), which indicates that the lattice strain is a major source of the transient creep strain.

36 MATERIALS SCIENCE↗

DOE OSTI2019

FY19 Status Report on Creep Test Data on Commercial Heat of Alloy 790

This report provides an update on the status of the creep rupture testing on the first commercial heat of Alloy 709 in support of the Alloy 709 Code Case development. The report is a Level 3 deliverable in FY19 (M3AT-19AN020502041), under the Work Package AT19AN0205204, “A709 Code Case Testing” performed by Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program. A creep test laboratory has been set up at ANL that comprises of 15 indirectly loaded test frames capable of testing ASTM standard-size specimens of the commercial heat of Alloy 709. The systems are equipped with high temperature furnaces capable of operation up to 1000°C in an air environment. In FY19, 15 creep-rupture tests of the ESR-1100°C solution annealed material, requiring the use of Type K thermocouples, were initiated to support the 100,000-hour Code Case. At the time of writing this report, eight tests were completed. This report presents the data from those completed tests and also microstructural characterization of the fractured specimens. During FY19, efforts were being made to add Type S thermocouples to five of the creep test frames, for performing longer term tests at higher temperatures.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Initial microstructural model for creep-fatigue damage in Grade 91 steel

This report describes an initial Crystal Plasticity Finite Element Method (CPFEM) model for cyclic plasticity and damage in Grade 91 steel. The objective of this work is to develop a framework for modeling creep-fatigue interaction in Grade 91 steel to better predict the onset of damage in high temperature microreactor components. Many microreactor concepts envision low operating pressures but relatively high thermal stresses. Under these conditions, creep-fatigue will likely be the dominant design failure mechanism. Physically based models, like the one under development here, could lead to a better understanding of creep-fatigue mechanisms and the effect of stress multiaxiality and hold time on creep-fatigue damage. In turn, this could lead to more efficient microreactor component designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

FY21 Status Report on Creep Rupture Testing at ANL to Support the Development of Alloy 709 Code Case

This report provides an update on the status of the creep rupture testing on the first commercial heat of Alloy 709 in support of the Alloy 709 Code Case development. Creep tests were performed in air on the lever-arm load frames in the Mechanical and Environmental Testing Laboratory at Argonne National Laboratory (ANL). In FY21, the focus was on the age-hardened ESR-1150°C material. Twenty two tests were initiated, among which ten were completed. This report presents the data from the completed tests. Metallographic characterization of the fractured specimens were performed and the results are included. The creep data enabled the preliminary analysis of the creep behavior of this material. The characterization showed the effect of test temperature and stress on the deformation microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Mechanistic Modeling of Creep and Aging for Time Extrapolation in LPBF 316 SS - LANL

A polycrystal constitutive model was extended to quantify the effects heterogeneity in the microstructure, porosity, and internal stresses on the mechanical response (creep, tensile) of 316H steel processed by additive manufacturing. The starting point is a recently developed and fully calibrated advanced microstructuresensitive mechanistic constitutive model fit to literature data pertaining to conventionally processed steels. The model is embedded within the full field fast Fourier transform framework (LApx) and several hundred simulations were performed. Synthetic microstructures with different grain aspect ratios were generated using the Neper package to study the role of grain morphology on the mechanical responses. The effect of initial cell interior dislocation density, cell-wall density, precipitate content, porosity, and internal stresses on creep rates are quantified for different polycrystalline microstructures. The study allowed to prospectively quantify the role of distinct features of the microstructure on the mechanical response under tension and creep loading scenarios. Critically it is found that minor anisotropies in the tensile response of samples, emanating from the aspect ratio of grains correlated into significant differences in creep rate as a function of loading directions. The study also allows to identify initial microstructure porosities below which the materials response only moderately degraded by the initial microstructure.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Microstructural evaluation of the creep behavior in L-PBF Ni-based superalloys

This presentation at ICAM 2024 Conference focuses on the commonalities and differences in the creep rupture behavior and creep mechanisms for three distinct classes of laser powder-bed fusion (L-PBF) Ni-based superalloys (γ’-precipitate strengthened Haynes® 282®, γ’/γ”/δ-precipitate strengthened Alloy 718, and solid-solution strengthened Alloy 625) as compared to conventionally processed counterparts. A comparison of Larson-Miller parameter plots establishes that these alloys perform statistically within the bounds established for the wrought product, despite having dissimilar microstructural features and other artefacts associated with PBF-LB manufacturing and post-processing heat treatment. To understand the failure and the impact of composition, minor phases, and deformation defects on creep behavior, the fractography has been performed and microstructures have been evaluated in detail with SEM-EDS, EBSD, and HAADF-STEM. The underlying diffusional and dislocation creep mechanisms associated with this microstructural evaluation is discussed. This work is supported by NETL-FWP-1022408 Advanced Turbines.

Sudbrack, Chantal↗

DOE OSTI2024

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Measure the effect of molten halide salt exposure on creep rupture lifetime

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Primary and Low-Strain Creep Models for 9Cr Tempered Martensitic Steels Including the Effects of Irradiation Softening and High-Helium Re-Hardening

Primary and low-strain creep represents a very important integrity challenge to large, complex structures, like fusion reactors. Here, we develop a predictive empirical primary creep model for 9Cr tempered martensitic steels (TMS), relating the applied stress (σ) to strain (ε), time (t) and temperature (T). The most accurate model is based on the applied σ normalized by the steel’s T-dependent ultimate tensile stress (σo), σ/σo(T). The model, fit to 17 heats of 9Cr TMS, yielded a σ root mean square error (RMSE) of ≈±11 MPa. Notably, the model also provides robust predictions for all the other TMS, when calibrated only by the fusion candidate Eurofer97 database. The model was extended to explore two possible effects of neutron irradiation, which produces both displacements per atom (dpa) and helium (He in atomic parts per million, appm) damage. These effects, which have not been previously considered, include: (a) softening, as a function of dpa, at T > ≈400–450 °C, in low-He fission environments (<1 He/dpa); and (b) subsequent re-hardening in high-He (≥10 He/dpa) fusion first-wall environments. The irradiation effect models predict (a) accelerated primary creep due to irradiation softening; and (b) fully arrested creep due to high-He re-hardening.

Alam, Md Ershadul (ORCID:0000000345968026)↗

DOE OSTI2017

Hierarchical creep cavity formation in an ultramylonite and implications for phase mixing

Abstract. Establishing models for the formation of well-mixed polyphase domains in ultramylonites is difficult because the effects of large strains and thermo-hydro-chemo-mechanical feedbacks can obscure the transient phenomena that may be responsible for domain production. We use scanning electron microscopy and nanotomography to offer critical insights into how the microstructure of a highly deformed quartzo-feldspathic ultramylonite evolved. The dispersal of monomineralic quartz domains in the ultramylonite is interpreted to be the result of the emergence of synkinematic pores, called creep cavities. The cavities can be considered the product of two distinct mechanisms that formed hierarchically: Zener–Stroh cracking and viscous grain-boundary sliding. In initially thick and coherent quartz ribbons deforming by grain-size-insensitive creep, cavities were generated by the Zener–Stroh mechanism on grain boundaries aligned with the YZ plane of finite strain. The opening of creep cavities promoted the ingress of fluids to sites of low stress. The local addition of a fluid lowered the adhesion and cohesion of grain boundaries and promoted viscous grain-boundary sliding. With the increased contribution of viscous grain-boundary sliding, a second population of cavities formed to accommodate strain incompatibilities. Ultimately, the emergence of creep cavities is interpreted to be responsible for the transition of quartz domains from a grain-size-insensitive to a grain-size-sensitive rheology.

Gilgannon, James (ORCID:0000000325971762)↗