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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Initial framework for engineering-scale statistical creep-fatigue modeling

This report describes the integration of new solid and interface-cohesive mechanics systems into MOOSE. The purpose of these new systems is to support the ability of MOOSE to run full-field crystal plasticity finite element method simulations of key material processes in high temperature metallic materials. These simulations could be used to help accurately predict the performance of key high temperature structural materials in future advanced nuclear reactor components. Previous work implemented preliminary versions of many of these systems in MOOSE Apps. The current work reports on their integration into the main MOOSE tensor mechanics module along with associated improvements to the basic formulations and numerical implementations. Finally, the report provides an example of the full-field crystal plasticity simulations now possible in MOOSE, including examples of realistic geometries requiring millions of degrees of freedom to resolve the microstructural features and macroscale geometry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Specimen sizing and remaining life sample calculations for high temperature reactor material surveillance

Advanced high-temperature nuclear reactors rely on structural components that will operate for decades under combined mechanical, thermal, and environmental loading. Materials surveillance programs are a promising strategy for managing the resulting uncertainty in long-term structural integrity by monitoring degradation in service using passively actuated mechanical test articles. Previous reports have developed a simplified, spreadsheetimplementable framework for sizing these test articles and for inferring accumulated creep damage and remaining life from ex-situ test data. This report advances that work toward practical deployment by providing a sample-problem book: a collection of worked, end-toend examples in which the ASME Section III, Division 5 design analysis of a representative high-temperature reactor component is carried through to a sized surveillance article, verified with detailed finite-element analysis, and concluded with a remaining-life assessment based on an assumed ex-situ creep-rate measurement on the retrieved specimen. The report also summarizes ongoing ANL engagement with ASTM Committee E10 on Nuclear Technology and Applications toward drafting a standard covering surveillance procedures for advanced reactors.

Barua, Bipul (ORCID:0000000247184113)↗
DOE OSTI2017

Carbon Contamination During Ion Irradiation - Accurate Detection and Characterization of its Effect on Microstructure of Ferritic/Martensitic Steels

Accelerator-based ion beam irradiation techniques have been used to study radiation effects in materials for decades. Although carbon contamination induced by ion beams in target materials is a well-known issue in some material systems, it has not been fully characterized nor quantified for studies in ferritic/martensitic (F/M) steels that are candidate materials for applications such as core structural components in advanced nuclear reactors. It is an especially important issue for this class of material because of the strong effect of carbon level on precipitate formation. In this paper, the ability to quantify carbon contamination using three common techniques, namely time-of-flight secondary ion mass spectroscopy (ToF-SIMS), atom probe tomography (APT), and transmission electron microscopy (TEM) is compared. Their effectiveness and shortcomings in determining carbon contamination are presented and discussed. The corresponding microstructural changes related to carbon contamination in ion irradiated F/M steels are also presented and briefly discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Crack Growth Rate and Fracture Toughness Tests on Irradiated Ex-Plant Materials

The performance of structural materials is critical for the safe and economic operation of light water reactors. Exposed to neutron irradiation during service, the reactor core internal materials can undergo significant microstructural and microchemical changes, leading to irradiation hardening and embrittlement. To ensure the structural integrity and functionality of nuclear reactor components during long-term operation, material degradation and damage mechanisms must be understood and managed adequately. In this work, irradiated materials harvested from the decommissioned Zorita reactor were studied for their cracking susceptibility and fracture resistance as a function of irradiation dose up to 47 displacement per atom (dpa). The material is a Type 304 stainless steel sectioned from the baffle plates of this pressurized water reactor with 38 years of service. Crack growth rate and fracture toughness J-resistance (J-R) curve tests were performed in low-corrosion-potential environments at ~315°C. All samples behaved similarly under cyclic loading, and no deteriorated corrosion-fatigue behavior was observed in the test environments. Under constant stress intensity factors, most samples did not show elevated crack growth rates, suggesting an adequate stress corrosion cracking resistance in the test environments. However, an unstable cracking behavior was observed in a 47-dpa sample, resulting in significantly higher crack growth rates than expected at high stress intensity factors. Crack instability was also observed in a 0.06-dpa sample but did not lead to a sustained high crack growth rate. The impact of neutron irradiation was more evident in the fracture toughness J-R curve tests. As the dose increased, the J-R curve declined considerably and became very shallow at high doses. A fully intergranular fracture morphology was also observed among the high-dose samples ruptured in an air atmosphere at room temperature. This brittle fracture mode in the absence of high temperature water environment confirmed a high degree of embrittlement of this material resulting from its service exposure to neutron irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Additive Manufacturing of Research Reactor Control Elements and Subsequent Neutron Irradiation

Additive manufacturing techniques have made major strides in recent years and are now capable of creating complex components in near-final form. The nuclear industry has many applications with complicated and even unique components that are time consuming, expensive, and more difficult to make using traditional fabrication routes. Modern fabrication techniques possess the potential to reduce fabrication lead-time and costs while increasing quality and efficiency. However, these methods have not been adopted due to the need for strict validation for nuclear applications. Unlike structural components that largely consist of a uniform material, for nuclear reactor core components incorporation of fissile or neutron poison materials inside a cladding becomes necessary. This in turn requires a more complex and flexible manufacturing method to produce these hybrid components. In this study, demonstration of ultrasonic additive manufacturing (UAM) process to produce components resembling research reactor control elements was carried out]. Specifically, the control elements for ORNL's High Flux Isotope Reactor (HFIR) were targeted. These control elements (consisting of an inner cylinder and four outer plates) are placed on the periphery of the circular HFIR core. The control elements consist of Al6061-T6 alloy with a 13-cm-long grey and a 56-cm-long black region, corresponding to 40 wt% of Ta and 30 wt% Eu{sub 2}O{sub 3} (europia) are incorporated along their length. In this study various fabrication parameters for production of aluminum blocks with and without embedded Eu{sub 2}O{sub 3} and Ta bearing compacts were explored. Select optimized specimens we then irradiated inside HFIR for one reactor cycle. The UAM process exploits ultrasonic energy to facilitate solid-state welding of metal foils in contact under pressure. This additive process is carried out in successive layers to build up the material and when coupled with CNC milling, it yields complex 3D components. The milling and subsequent additive steps may be used to embed foreign material into the material undergoing buildup. In this manner fissile or neutron poison elements may be incorporated into the complex structure, though in an inhomogeneous fashion. Multiple Al-6061 blocks were produced by UAM using 25 mm wide by 0.15 mm thick Al-6061-H18 tapes. The mechanical properties of these blocks in the as-fabricated condition was characterized in detail and it was shown that a great degree of anisotropy in the mechanical behavior exists between the direction in the tape plane and the direction perpendicular to the tapes (buildup direction) The latter exhibited reduced modulus and significantly smaller strength. Post production heat treatment appeared to greatly enhance the strength in the buildup direction. Multiple tensile specimens machined per SS-J3 geometry were prepared for irradiation testing. The tensile specimen matrix was set up to enable post irradiation examination of mechanical properties in various directions as well as after heat treatment. A small block with embedded Ta- and Eu{sub 2}O{sub 3}-bearing compacts, three of each kind, was also produced, as demonstrated previously. These compacts were 5.7 mm in diameter and 1.6 mm in height and contained 60 vol Ta or Eu{sub 2}O{sub 3}. The poison-bearing Al-6061 block alongside the tensile bars was loaded into a perforated rabbit (irradiation capsule) for irradiation in HFTK. The perforated rabbit allows penetration of HFIR water coolant into the Interior of the capsule in direct contact with the test specimens. The rabbit was loaded into the flux trap region at the uppermost position in the core. Irradiation temperature was 68 deg. C and fast neutron (E>0.1 MeV) fluence was 10{sup 25} neutrons/m{sup 2}. Currently PIE activities are ongoing that are focused on mechanical testing of irradiated tensile bars and metallographic examination of neutron poison bearing block.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Molten Salt Reactors and Electrochemical Reprocessing: Synthesis and Chemical Durability of Potential Waste Forms for Metal and Salt Waste Streams

The molten salt reactor (MSR) is one of the leading advanced nuclear reactor candidates to replace current nuclear reactor technologies in the U.S. Besides having more economical and reliable designs, MSRs have are amenable to a closed fuel cycle, in which electrochemical reprocessing can be performed to recycle the used nuclear fuel. This review intends to provide information about potential waste forms for metal and salt waste streams from these salt-based nuclear processes. Metal waste streams arise from reactor components and structural materials. Salt waste streams are generated during reactor operations as fission products build up in salt-fueled systems. Waste forms that have the highest waste loading and/or have shown the most commercial promise are discussed with an emphasis on the current state of efforts to understand the synthesis and chemical durability of metal and ceramic waste forms.

molten salt reactors, electrochemical reprocessing↗
DOE OSTI2022

Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components (Final Technical Report)

Metal Additive Manufacturing (AM) is a promising method for cost-efficient fabrication of complex shape structures for applications in harsh environment, such as in a nuclear reactor. However, internal defects (pores) occur in high-strength AM alloys, which are manufactured with Laser Powder Bed Fusion (LPBF) AM method. Pulsed Infrared Thermography (PIT) is an efficient nondestructive evaluation (NDE) method to examine actual structures, because this method offers one-sided non-contact measurements, and fast processing of large sample areas. However, imaging of material defects, particularly defects with sizes at microscopic level, is challenging. In this report, we benchmark the performance of several Unsupervised Learning (UL) algorithms designed to enhance imaging of microscopic defects in metals with PIT. UL aims to learn the latent principal patterns (dictionaries) in PIT data to detect defects with minimal human supervision. Performance of Independent Component Analysis (ICA), Sparse Coding (SC), Principal Component Analysis (PCA) and Exploratory Factor Analysis (EFA) was compared using F-score, UL model training time and defects reconstruction time. We obtained the average F-score of 0.75, and a highest F-score of 0.89 for the EFA algorithm. Overall, EFA outperforms other UL algorithms considered in this study. In another approach, we investigate Thermal Tomography (TT), which is a computational method for reconstruction of depth profile of internal material defects from PIT nondestructive evaluation (NDE). TT algorithm obtains depth reconstructions of thermal effusivity, which has been shown to provide visualization of subsurface internals defects in metals. In many applications, one needs to determine the defect shape and orientation from reconstructed effusivity images. Interpretation of TT images is non-trivial because of blurring, which increases with depth due to heat diffusion-based nature of image formation. We have developed a deep learning convolutional neural network (CNN) to classify size and orientation of subsurface material defects in TT images. CNN was trained with TT images produced with computer simulations of 2D metallic structures (thin plates) containing elliptical subsurface voids. Performance of CNN was investigated using test TT images developed with computer simulations of plates containing elliptical defects, and defects with shape imported from scanning electron microscopy (SEM) images. CNN demonstrated the ability to classify radii and angular orientation of elliptical defects in previously unseen test TT images. We have also demonstrated that CNN trained on TT images of elliptical defects is capable of classifying shape and orientation of irregular defects. Training the CNN on irregular defect shapes instead of on elliptical shapes would make the resulting classifications more descriptive of actual defect shapes. However, this requires a much higher volume of SEM images of material defects, which are difficult to obtain because of random occurrence of defects in LPBF. To address this challenge, we developed a generative adversarial network (GAN) to augment the existing dataset of SEM defect images. The GAN model is demonstrated to create novel yet realistic defect shapes that can be used as input for simulated PTT images to train CNN. We also investigate several approaches based on Gaussian Random Circle and Bezier Curves for constructing parametric models of irregular-shape defects.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Lach, Tim↗
DOE OSTI2023

Mechanical Test Results and Microstructural Characterization of the Harvested and Baseline Archival Zion RPV Materials

The decommissioning of the Zion Units 1 and 2 Nuclear Generating Station in Zion, Illinois, presented a unique opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Office of Nuclear Energy, Light Water Reactor Sustainability (LWRS) Program, coordinated with Zion Solutions, LLC, a subsidiary of Energy Solutions (ES), the selective procurement of materials, structures, and components, from the decommissioned reactors including multiple segments of the Zion Unit 1 Reactor Pressure Vessel (RPV).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Neutron Irradiation of Nuclear Structural Materials Fabricated by Powder Metallurgy with Hot Isostatic Pressing

An irradiation campaign intended to produce data necessary to qualify various reactor structural materials fabricated by powder metallurgy with hot isostatic pressing (PM/HIP) is currently in progress. Manufacturing processes have considerable influence over the safety and integrity of nuclear reactor vessels and internal components. Established processes such as casting, plate rolling and welding, forging, drawing, and extrusion have been used to fabricate structural and pressure-retaining materials used in the nuclear power industry for the past 60 years. However, weldability and inspectability are ongoing issues in the manufacture of reactor vessels and internals. Furthermore, these components are subject to harsh service environments, which combine high radiation fluence, high temperature, and mechanical stress. These conditions can accelerate material degradation and failure, with the most extreme degradation often occurring at weldments and in poorly manufactured or inspected components. Advanced reactor designs and lifetime extensions to the existing fleet of light-water reactors (LWRs) will exacerbate materials degradation issues by increasing the duty on reactor internals. Thus, developing reliable manufacturing processes wherein high-quality weldments and inspections can be performed is of great importance to the continued safety and operation of nuclear power plants.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

MACHINING TEST SPECIMENS FROM HARVESTED ZION RPV SEGMENTS

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials, structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].

Reactor Pressure Vessel↗
DOE OSTI2019

Chapter 4 - Research Tools: Microstructure, Mechanical Properties, and Computational Thermodynamics

The purpose of this Chapter is to provide an overview of the research tools that are available to evaluate the response of materials that are subject to atomic displacements caused by fast particle irradiation. The major focus is on metals employed in structural components in operating or planned nuclear fission and fusion reactors. Understanding, predicting and characterising the changes in material properties under irradiation requires the ability to characterise the irradiation-induced microstructure at and above the atomic scale, and the ability to relate such changes to the observed degradation in performance sustaining properties. Thus, any discussion of research tools necessarily involves microstructural characterization techniques and mechanical property characterization techniques. In addition, computational thermodynamics (and related multiscale, multiphysics modeling) can also be considered characterization tools since they enable the study of the thermodynamically and kinetically mediated mechanisms (influenced by the irradiation environment) that affect microstructural evolution. Hence, this chapter also contains a brief review of computational thermodynamics.

English, Colin A.↗
DOE OSTI2021

Development of Real-Time, In-pile Creep Test Rigs for Characterizing the Structural Materials of Nuclear Components

New and improved materials are being considered for supporting both existing and next-generation nuclear reactors. Reactor materials can significantly degrade with time, thus limiting or altering their properties in harsh reactor environments. To accurately understand such material degradation, real-time data obtained under prototypic irradiation conditions are required. In particular, understanding the creep behavior of materials exposed to irradiation and elevated temperatures is essential for safety concern evaluations. To provide these capabilities, Idaho National Laboratory (INL)’s High Temperature Test Laboratory (HTTL) developed several instrumented test rigs for obtaining real-time data from specimens in well-controlled pressurized-water reactor (PWR) coolant conditions at the Materials Test Reactor. This technical report focuses on INL’s efforts to evaluate and enhance the former creep test rig prototype that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig can detect changes in the length of creep specimens, which is useful for measuring thermal expansion and creep deformation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Summary of Structural Material Modeling Development for the NEAMS Program in Fiscal Year 2020

This report summarizes work performed during Fiscal Year (FY) 2020 at Idaho National Laboratory (INL) for the U.S. Department of Energy?s Nuclear Engineering Advanced Modeling and Simulation (NEAMS) program for the Structural Materials and Chemistry Technical Area in the work package entitled "MS- 20IN050104 - Structural Materials - INL." The Structural Materials and Chemistry Technical Area is a relatively new component of the NEAMS program, and is currently focusing on developing simulation capabilities to support the deployment of nuclear energy in the areas of molten salt reactor chemistry, light water reactor (LWR) structural material degradation, and structural material behavior for advanced reactor applications. INL performed work for to advance capabilities for simulation of structural material behavior in both LWR and advanced reactor applications in the work described here.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Sensitivity Study of Multiscale and Phenomenological Elasto-Viscoplastic Grade 91 Material Models for Component-Scale Response

Many advanced nuclear reactor concepts currently being developed are targeting higher operating temperatures relative to the current fleet of light water nuclear reactors, for efficiency gains and other operational considerations. The design of high temperature structural components with reliable long-term operational performance will depend on material models that accurately capture the inelastic deformation mechanisms active in these environments. In this work, we perform a detailed parameter sensitivity analysis of two unified elasto-viscoplastic Grade 91 material models capable of capturing long term high temperature creep deformation. The first model is a phenomelogical material model from the Nuclear Engineering Material Library (NEML) developed at Argonne National Lab. The NEML model parameters and their uncertainty were fit to a range of Grade 91 experimental data using Bayesian Markov Chain Monte Carlo analysis. The second model is a LAROMance data-driven surrogate material model developed at Los Alamos National Lab. The LAROMance model is fit to a large database of responses produced by a mechanistic crystal plasticity based polycrystal model. Parameters for the LAROMance surrogate material model reflect the pedigree of the Grade 91 microstructure. Both material models have been integrated into the Grizzly code, based on the open-source MOOSE multiphysics simulation framework, to simulate both the progression of aging mechanisms and the effects of that aging on nuclear power plant structures. Grizzly is used analyze a three-dimensional Grade 91 piping system to compare the long-term inelastic response predicted by these two fundamentally different models and assess the sensitivity of the material model input parameters on this quantity of interest.

42 ENGINEERING↗
DOE OSTI2020

Design and strategy for next-generation silicon carbide composites for nuclear energy

Silicon carbide (SiC) ceramic-based composites continue to be attractive material options for fusion in-vessel components and fission reactor core structures because of their exceptional high-temperature capability and favorable neutronic properties. As performance data accumulates, the limitations of the current generation of nuclear-grade SiC composites are becoming more apparent. These limitations mainly involve strength degradation during high-dose neutron irradiation. In this paper, we discuss several options for improving the performance of the next generation of SiC composites to enhance the radiation resistance, along with new experimental results on neutron irradiation resistance. The main emphasis is on the selection of the fibers and the design and development of alternative interphase layers for advanced composites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗