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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

The mechanical interaction of clad and base metal for molten salt reactor structural components

Cladding structural components with a corrosion resistant material may greatly extend the design life of molten salt reactor concepts. A complete design methodology for such cladded, high temperature nuclear components will require addressing many issues: fabrication, corrosion resistance, metallurgical interaction, and the mechanical interaction of the clad and base materials under load. This work focuses on the final issue: the mechanical interaction of the base and clad under creep-fatigue conditions. Depending on the relative mechanical properties of the two materials the clad may substantially influence the longterm cyclic response of the structural system or its effect might be negligible. To quantify the effect of different clad material properties we develop an efficient method for simulating pressurized cladded components in the limiting case where the section of interest is far from structural discontinuities. Using this method we evaluate the mechanics of the clad/base system and identify different regimes of mechanical response. The focus is on situations relevant to high temperature nuclear components: thermal-cyclic Bree-type problems and similar axisymmetric structures. The insights gained from these structural studies will form the basis for developing design rules for high-temperature, nuclear, cladded components.

Messner, M. C.↗
DOE OSTI2021

Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components. Third Annual Progress Report

Additive manufacturing (AM) of high-strength corrosion resistance alloys for nuclear energy applications, such as stainless steel and Inconel, is currently based on laser powder bed fusion (LPBF) process. Some of the challenges with using LPBF method for nuclear manufacturing include the possibility of introducing pores into metallic structures. Probability of crack initiation at the pore depends on size, shape, and orientation of the defect. Pulsed Infrared Thermography Imaging (PIT) provides a capability for non-destructive evaluation (NDE) of sub-surface defects in arbitrary size structures. The PIT method is based on recording material surface temperature transients with infrared (IR) camera following thermal pulse delivered on material surface with flash light. The PIT method has advantages for NDE of actual AM structures because the method involves one-sided non-contact measurements and fast processing of large sample areas captured in one image. Following initial qualification of an AM component for deployment in a nuclear reactor, a PIT system can also be used for in-service nondestructive evaluation (NDE) applications. In this report, we describe recent progress in enhancing PIT capabilities in detecting microscopic subsurface defects in metals, and classifying shapes and orientation of pores in thermal images. For detection of microscopic defects in PIT imaging data, we have developed Spatial Temporal Denoised Thermal Source Separation (STDTSS) unsupervised machine learning (ML) image processing algorithm. We show that flat bottom hole (FBH) defects as small as 200µm in SS316 and IN718 specimens, can be detected with STDTSS algorithm. To the best of our knowledge, these are the smallest detected defects which are reported in literature. For classification of defects shapes, we have previously developed thermal tomography (TT) algorithm to obtain depth reconstructions of material defects from data cube of sequentially recorded surface temperatures. However, interpretation of TT images is non-trivial because of blurring with increasing depth. To address this challenge, we have developed a deep learning convolutional neural network (CNN) to classify size and orientation subsurface defects in simulated TT images.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Virtual ultrasound sources for inspecting nuclear components of coarse-grained structure

This work describes an ultrasonic inspection procedure designed for verifying coarse-grained structure materials, which are commonly used on nuclear reactors. In this case, conventional phased array techniques cannot be used due to attenuating characteristics and backscattered noise from microstructures inside the material. Thus, synthetic aperture ultrasonic imaging (SAFT) is used for this approach in contact conditions. In order to increase energy transferred to the medium, synthetic transmit aperture is formed by several elements which generate a diverging wavefront equivalent to a virtual ultrasound source behind the transducer. On the other hand, the phase coherence technique has been applied to reduce more structural noise and improve the image quality. The beamforming process has been implemented over a GPU platform to reduce computing time.

42 ENGINEERING↗
DOE OSTI2013

Investigation of a Novel NDE Method for Monitoring Thermomechanical Damage and Microstructure Evolution in Ferritic-Martensitic Steels for Generation IV Nuclear Energy Systems

The main goal of the proposed project is the development of validated nondestructive evaluation (NDE) techniques for in situ monitoring of ferritic-martensitic steels like Grade 91 9Cr-1Mo, which are candidate materials for Generation IV nuclear energy structural components operating at temperatures up to ≈650 °C and for steam-generator tubing for sodium-cooled fast reactors. Full assessment of thermomechanical damage requires a clear separation between thermally activated microstructural evolution and creep damage caused by simultaneous mechanical stress. Creep damage can be classified as "negligible" creep without significant plastic strain and "ordinary" creep of the primary, secondary, and tertiary kind that is accompanied by significant plastic deformation and/or cavity nucleation and growth. Under negligible creep conditions of interest in this project, minimal or no plastic strain occurs, and the accumulation of creep damage does not significantly reduce the fatigue life of a structural component so that low-temperature design rules, such as the ASME Section III, Subsection NB, can be applied with confidence. The proposed research project will utilize a multifaceted approach in which the feasibility of electrical conductivity and thermo-electric monitoring methods is researched and coupled with detailed post-thermal/creep exposure characterization of microstructural changes and damage processes using state-of-the-art electron microscopy techniques, with the aim of establishing the most effective nondestructive materials evaluation technique for particular degradation modes in high-temperature alloys that are candidates for use in the Next Generation Nuclear Plant (NGNP) as well as providing the necessary mechanism-based underpinnings for relating the two. Only techniques suitable for practical application in situ will be considered. As the project evolves and results accumulate, we will also study the use of this technique for monitoring other GEN IV materials. Through the results obtained from this integrated materials behavior and NDE study, new insight will be gained into the best nondestructive creep and microstructure monitoring methods for the particular mechanisms identified in these materials. The proposed project includes collaboration with a national laboratory partner and the results will also serve as a foundation to guide the efforts of scientists in the DOE laboratory, university, and industrial communities concerned with the technological challenges of monitoring creep and microstructural evolution in materials planned to be used in Generation IV Nuclear Energy Systems.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewhere since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en , which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Microstructural and material property changes in severely deformed Eurofer-97

Severe plastic deformation changes the microstructure and properties of steels, which may be favourable for their use in structural components of nuclear reactors. In this study, high-pressure torsion (HPT) was used to refine the grain structure of Eurofer-97, a ferritic/martensitic steel. Electron microscopy and X-ray diffraction were used to characterise the microstructural changes. Following HPT at room temperature to a maximum shear strain of 230, the average grain size reduced by a factor of ~30, with a marked increase in high-angle grain boundaries. Dislocation density also increased by more than one order of magnitude. The thermal stability of the deformed material was investigated via in-situ annealing during synchrotron X-ray diffraction. This revealed substantial recovery between 450 K – 800 K. Irradiation with 20 MeV Fe-ions to ~0.1 dpa caused a 20% reduction in dislocation density compared to the as-deformed material. However, HPT deformation prior to irradiation only had a minor effect in mitigating the irradiation-induced reductions in thermal diffusivity and surface acoustic wave velocity of the material. Microstructural and material property changes are dominated by deformation compared to irradiation. In light of this, the benefits of using HPT to improve the irradiation resistance of Eurofer-97 are limited. These results provide a multi-faceted view of the changes in ferritic/martensitic steels due to severe plastic deformation, and how these changes can be used to alter material properties.

ion-irradiation↗
DOE OSTI2012

Presentation of FP7 matter project: general overview

The 2010-2012 implementation plan of the European Sustainable Nuclear Industrial Initiative (ESNII), prepared in the frame of the Sustainable Nuclear Energy Technology Platform (SNETP), establishes a very tight time schedule for the start of construction of the European Gen IV prototypes; namely the construction of the LFR ETPP (European Technology Pilot Plant) MYRRHA will start in 2014 and that of the SFR Prototype ASTRID will start in 2017. The GEN IV reactors pose new challenges to the designers and scientists in terms of higher operating temperature, higher fuel burn-up, and in some cases more corrosive environment with respect to the present technologies and which impacts the materials performance. In this frame, the MATTER (Materials Testing and Rules) Project starts well targeted R and D activities to perform careful materials studies in GEN IV operational conditions and to find out criteria for the correct use of these materials in relevant reactor applications. Aim of the MATTER Project (that involved 27 partners and will end in 2015) is to complement the materials researches, in the frame of the European Energy Research Alliance (EERA) guidelines, with the implementation of pre-normative rules. The MATTER Project is divided in 3 technical Domains (called DM): DM1 - Development of test and evaluation guidelines for structural materials: to develop/establish best practice guidelines for testing and evaluation procedures, which are aimed to screen and characterize nuclear materials for innovative nuclear systems. DM2 - Pre-normative R and D for Codes and Standards: Pre-normative activities are performed, comprehensive of experiments, to revise and update the design rules (with an EU level consensus) in order to answer to some short term needs of the two projects ASTRID and MYRRHA with respect to the design and the construction of structural components. DM3 - Joint Program Scheme, implementation and Priorities: to optimise the effectiveness and efficiency of the EERA Joint Program on nuclear materials for innovative reactors and to support specific research activities related to fundamental understanding of ODS steels fabrication. ODS steels are considered candidate materials, in the medium-long term, for high fuel burn-up cladding application. After a brief presentation of DM1 and DM3, this paper mainly focuses on description of Pre-normative R and D activities for Codes and Standards (DM2). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Determining the Effects of Neutron Irradiation on the Structural Integrity of Additively Manufactured Heat Exchangers for Very Small Modular Reactor Applications, DOE Final Report (Project # 19-16980)

Auburn University (AU) teamed with the University of Missouri Research Reactor (MURR) and Kansas State University (KSU) to determine how to best use laser-powder bed fusion (L-PBF) additive manufacturing (AM) methods for generating radiation resistant nickel-based superalloys, Inconel alloy 625 and 718, for special purpose reactor (SPR) or very small modular reactor (vSMR) heat-exchangers (HeXs). Compact, conformal, and durable HeXs that are tolerant of extreme environments are needed for supporting the technical maturity of next-generation, portable compact reactors. AM is an enabler for realizing this new wave of HeXs – providing a means to make customizable hot and cold stream architectures with novel flow path geometries (e.g., tortuous channels with non-uniform, asymmetric cross-sections) and reduced layer-to-layer contact resistance (i.e., no separate bonding procedure required). AM further enables a more time/cost efficient means for fabricating SPRs by reducing the number of suppliers required for HeX assembly and allowing for on-site HeX fabrication. The project aim has been to better understand how neutron irradiation affects the microstructure and properties of additively manufactured nickel-based superalloys, to accelerate their safe, reliable use in the modular reactor industry. The major objective was to qualify/quantify the microstructure and microhardness of nickel-based superalloys (including Inconel 718 and 625) additively manufactured via the L-PBF process in the neutron-dosed (irradiated) and non-irradiated states over a course of 3 years. Effects of build orientation during L-PBF and post-AM heat treatments on neutron resistance, microstructure and mechanical properties were also investigated. Neutron damage mechanisms via hardening were characterized. This project combined subject-matter experts in AM, mechanical/microstructure metallic part characterization, and neutron irradiation, as well as unique assets and capabilities at AU and MURR at MU, to ensure project results translated to effectively addressing known gaps in nuclear science and engineering. Parts were fabricated using L-PBF systems readily available at AU. Specimens were then irradiated using MURR facilities; a manipulator equipped hot cell was also used to measure material hardness after dosing. MURR, a 10 MW, light-water nuclear reactor, is the largest, highest powered, highest-flux university owned research reactor in the U.S. The major findings in this project provide evidence that AM can serve as an alternative way to build structural components for future advanced small modular reactors using advanced materials like Inconel 625 (IN625) or Inconel 718 (IN718). After full spectrum neutron irradiation, vertically as-built AM IN625 samples were observed to display better resistance towards radiation-induced-hardening defects relative to traditionally machined metals. A Vickers microhardness tester, using settings of 1 kgf and dwell time of 15 seconds per indentation, was used to measure hardness in this study. The as-built, vertically printed samples experienced 1.2% of radiation hardening vs. 5.25% radiation hardening observed in wrought IN625. Another set of IN625 and IN718 samples were exposed to fast neutron irradiation. It was observed that IN718 showed more resistance towards radiation hardening compared to IN625 samples indicating IN718 had a better performance. Results showed that the IN718 samples experienced less change (-2.5 to 3.24 %) in microhardness. On the other hand, IN625 samples underwent more (0.9 to 7.21%) change in microhardness after fast neutron irradiation. AM IN625 samples were irradiated using an ion (proton) beam in cyclotron. The mechanical properties of AM samples post irradiation were compared with wrought samples. The irradiated region on the samples were tested using nano-hardness indention. It was observed that the beam current and time used in this study generated an annealing effect and thus reduced the hardness of the samples. The sum of the project results provide precious insight into how one may minimize radiation hardening in AM materials while maintaining material property constraints. Results should assist engineers in selecting an appropriate heat treatment for AM nickel-based superalloys for increased radiation resistance. Results should increase confidence levels for adopting AM for building nuclear reactor components which perform the same or better than conventionally manufactured components. Fast neutron irradiation testing provided an accelerated means of obtaining radiation effects without making materials radioactive and difficult to handle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Evaluation of Binder Jetting and Chemical Vapor Infiltration as a Production Method for Silicon Carbide Parts for Nuclear Applications

Application of silicon carbide (SiC) is often sought within nuclear energy systems due to its high thermal conductivity, mechanical and environmental stability at elevated temperatures, corrosion resistance, and exceptional displacement damage irradiation tolerance. As such, the transformational challenge reactor (TCR) program is considering its application in its additively manufactured core. SiC may be used as the matrix for the TCR fuel blocks that comprise of TRISO fuel particles embedded inside this material or used as core non-fuel-bearing peripherical structures such as bottom support manifolds. The core peripherical structures serves to support the overall core geometry, direct and guide coolant flow, and also reflect back leaked neutrons into the core. SiC has very low neutron absorption and adequate scattering cross sections that makes it an ideal material for utilization in the TCR core. Like many ceramics, however, SiC is difficult and at times impractical to shape into complex geometries via traditional machining due to its brittleness. Further, large structures comprising nuclear-grade SiC, i.e. highly pure and crystalline, are costly. Additive manufacturing (AM), however, offers the ability to shape materials by adding layer-by-layer rather than subtracting, contributing new possibilities for previously difficult-to-shape materials. Thus, exploring AM as a method to shape SiC offers a highly cost-effective method and is strategic to developing the next generation of components for nuclear reactors. This report outlines an evaluation of binder jet additive manufacturing combined with chemical vapor infiltration (CVI) to produce nearly-dense silicon carbide artifacts in complex geometries suitable for advanced nuclear reactor designs. The focus is placed on the binder jet printing technologies at large scale for nonfuel-bearing structures relevant to the TCR core and other advanced reactors. Among all AM processes, binder jet delivers the highest resolution capability while still maintaining large print sizes, making it the ideal candidate for nuclear component production.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Handbook of the Materials Properties of FeCrAl Alloys For Nuclear Power Production Applications

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including as accident tolerant fuel cladding, structural components for fast fission reactors, and as first wall and blanket structures for fusion reactors. FeCrAl alloys are under consideration for these applications due to their inherent corrosion resistance, stress corrosion cracking resistance, radiation-induced swelling resistance, and high temperature oxidation resistance. A substantial amount of research effort has been completed to design, develop, and begin commercial scaling of FeCrAl alloys for nuclear power applications over the past half a century. These efforts have led to the development of an extensive database on material properties and process knowledge for FeCrAl alloys but not within a consolidated format. The following report is the first edition of a materials handbook to consolidate the state-of-the-art on FeCrAl alloys for nuclear power applications. This centralized database focuses solely on wrought FeCrAl alloys, oxide dispersion strengthened alloys, although discussed in brief, are not covered. Where appropriate, recommendations for applications of the data is provided and current knowledge gaps are identified.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Structures, Systems, and Components Classification Criteria for the Versatile Test Reactor

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (DOE), Office of Nuclear Energy. The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by the U.S. National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc. As part of the VTR authorization process by the DOE, it is necessary to identify the safety classification of structures, systems, and components (SSCs) to determine the proper design, surveillance, operation, and incident-reporting requirements. Since VTR is utilizing a risk-informed, performance-based authorization process based on the Licensing Modernization Project (LMP) described in NEI 18-04 [1], the criteria for SSC classification, which are reviewed here, includes both risk information from the probabilistic risk assessment (PRA) and prescriptive requirements.

Grabaskas, David↗
DOE OSTI2022

Designing Cladded Components for High Temperature Nuclear Service: Part II—Design Rules

The challenge of using existing ASME Section III, Division 5, class A metallic materials for the construction of structural components of advanced reactors with corrosive coolants could be mitigated by allowing designers to use cladding to protect the base material from corrosion. However, the existing Section III, Division 5 rules provide no guidance on the evaluation of strain accumulation and creep-fatigue damage in cladded components. The availability of design rules for cladded components that do not require long-term clad material testing could promote the application of the cladding approach to accelerate the deployment schedule of these advanced reactor systems. To avoid long-term properties for the clad materials Part I of this work proposes two approximate design analysis methods for two types of clad materials—soft clads that creep much faster and have lower yield stress than the class A base material, and hard clads that creep much slower and have higher yield stress than the class A base material. The proposed analysis methods approximate the response of a soft clad by treating it as perfectly compliant and of a hard clad by treating it as linear elastic. Based on these approximate design analysis strategies this Part II develops a complete set of design rules for class A components cladded with either soft or hard clad materials. In conclusion, Part II discusses the reasoning behind the proposed design rules and uses example finite element analyses of representative reactor components to illustrate the use of these design methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Continuum Damage Mechanics Modeling of High-Temperature Flaw Propagation: Application to Creep Crack Growth In 316H Standardized Specimens and Nuclear Reactor Components

Predicting creep crack growth (CCG) of flaws found during operation in high-temperature alloy components is essential for assessing the remaining lifetime of those components. While defect assessment procedures are available for this purpose in design codes, these are limited in their range of applicability. This study assesses the application of a local damage-based finite-element methodology as a more general technique for the prediction of CCG at high temperatures on a variety of structural configurations. Numerical results for stainless steel 316H, which are validated against experimental data, show the promise of this approach. This integration of continuum damage mechanics (CDM) based methodologies, together with adequate inelastic models, into assessment procedures can therefore inform the characterization of CCG under complex operating conditions, while avoiding excessive conservatism. This article shows that such modeling frameworks can be calibrated to experimental data and used to demonstrate that the degree of tri-axiality ahead of a growing creep crack affects its rate of growth. The framework is also successfully employed in characterizing CCG in a realistic reactor pressure vessel geometry under an arbitrary loading condition. These results are particularly relevant to the nuclear power industry for defect assessment and inspections as part of codified practices of structural components with flaws in high-temperature reactors.

42 ENGINEERING↗
NASA NTRS2001

Creep-Fatigue Interaction Testing

Fatigue fives in metals are nominally time independent below 0.5 T(sub Melt). At higher temperatures, fatigue lives are altered due to time-dependent, thermally activated creep. Conversely, creep rates are altered by super. imposed fatigue loading. Creep and fatigue generally interact synergistically to reduce material lifetime. Their interaction, therefore, is of importance to structural durability of high-temperature structures such as nuclear reactors, reusable rocket engines, gas turbine engines, terrestrial steam turbines, pressure vessel and piping components, casting dies, molds for plastics, and pollution control devices. Safety and lifecycle costs force designers to quantify these interactions. Analytical and experimental approaches to creep-fatigue began in the era following World War II. In this article experimental and life prediction approaches are reviewed for assessing creep-fatigue interactions of metallic materials. Mechanistic models are also discussed briefly.

Halford, Gary R.↗
DOE OSTI2016

U.S. implementation of Q12{sup TM} zirconium alloy for improved stability of structural components

AREVA NP undertook the Q12{sup TM} research and development program to optimize the contribution to fuel assembly dimensional stability of structural components fabricated from ultra-low tin Zr-1Nb-Sn-Fe quaternary alloys. Because Q12{sup TM} is an evolutionary development of the current M5 alloy metallurgy, it benefits from the extensive industrial experience already gained with the latter. The Q12{sup TM} alloy is processed in the same way as M5, with slight increases in tin (Sn) and iron (Fe) contents. For structural components, these chemistry modifications provide higher irradiation creep strength and improved corrosion performance over Zircaloy-4. Fuel assemblies with quaternary alloy guide tubes are thus expected to increase robustness with respect to fuel assembly bow, improve growth characteristics, and maintain high burnup capability. The basic material properties have been determined through various in-core and out-of-core test programs. Components made from Q12{sup TM} have undergone irradiation in lead assemblies and in batch fuel for 13 Pressurized Water Reactor (PWR) units worldwide. Fuel with Q12{sup TM} guide tubes and grids has achieved a fuel assembly average burnup of 60 GWd/tU. Post irradiation examinations of these components have demonstrated that Q12{sup TM} performs in a stable, predictable manner. AREVA NP has submitted a topical report (TR) for Q12{sup TM} as a structural material to the United States Nuclear Regulatory Commission (US-NRC). This TR outlines the material behavior and performance of Q12{sup TM} for both guide tubes and grids. Once approved, implementation of Q12{sup TM} material throughout US PWR fuel designs will enhance dimensional stability of structural components. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

AGC-3 Specimen Post-Irradiation Examination Data Package Report

The Advanced Reactor Technologies Graphite research and development program is conducting an extensive graphite irradiation program to provide data to assist in licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.[ ] Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be a critical parameter in determining the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. A second, much more significant, internal stress is experienced within the graphite components due to irradiation induced dimensional changes. These induced stresses result from neutron irradiation causing physical changes in the graphite structure, first shrinking it and then expanding it, with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear graphite grades is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Handbook on the Material Properties of FeCrAl Alloys for Nuclear Power Production Applications (FY18 Version: Revision 1)

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including accident tolerant fuel cladding, structural components for fast fission reactors, and as first wall and blanket structures for fusion reactors. FeCrAl alloys are under consideration for these applications due to their inherent corrosion resistance, stress corrosion cracking resistance, radiation-induced swelling resistance, and high temperature oxidation resistance. A substantial amount of research effort has been completed to design, develop, and begin commercial scaling of FeCrAl alloys for nuclear power applications over the past half a century. These efforts have led to the development of an extensive database on material properties and process knowledge for FeCrAl alloys but not within a consolidated format. The following report is the first edition of a materials handbook to consolidate the current state of knowledge on FeCrAl alloys for nuclear power applications. This centralized database focuses solely on wrought FeCrAl alloys, oxide dispersion strengthened alloys, although discussed in brief, are not covered. Where appropriate, recommendations for applications of the data is provided and current knowledge gaps are identified.

36 MATERIALS SCIENCE↗
DOE OSTI2019

PTT System Performance Evaluation in 3-D Imaging of AM Components

Additive manufacturing (AM, or 3D printing) for commercial nuclear energy applications is an emerging method for cost-efficient manufacturing aimed at replacing aging nuclear reactor parts and reducing costs for new construction. Because of the geometry of metallic structures of interest for nuclear applications, which consist of planar primitives with no symmetry of revolution, limited options are available for non-destructive evaluation (NDE) either during or post manufacturing. Known material flaws in AM include low density regions consisting of nonsintered powder, which have to be detected to ensure the safety of long-term performance nuclear reactor components. As a solution to NDE of AM, we are developing pulsed thermal tomography (PTT) models and depth inversion algorithms for 3D imaging. PTT has many advantages because the method is non-contact and allows for in-service, NDE of AM nuclear reactor parts. By analyzing transients of surface temperature response due to internal thermal resistances, one can obtain 3D reconstructions of material effusivity using a unique inversion algorithm developed at Argonne. This study investigates the capability of PTT in imaging several AM structures fabricated from Inconel 718 (IN718) feedstock. The AM structures have complex geometry, but do not have calibrated internal defects. The objective is to determine PTT settings, such as the total integration time, for imaging of representative AM structures. A new FLIR X8500sc IR camera with higher spatial resolution (1280x1024 at 181Hz frame rate) was integrated into the PTT laboratory setup. Software tools were developed to ensure compatibility of new camera files with legacy effusivity reconstruction codes. The new camera was used to obtain 3-D reconstructions of IN718 nozzle plate and a particle filter plate. The results demonstrate the capability of PTT in 3- D imaging of complex geometry AM structures. Further, increased resolution imaging is achieved compared to prior results using older FLIR SC4000 IR camera. In addition, preliminary studies were conducted on developing image processing algorithms for automated detection of flaws. Initial demonstrations are shown for segmentation of PTT images of flat bottom hole (FBH) specimens. Imaging of FBH specimens was described in a prior report. Intermediate results demonstrated in this report will be developed further in detection of flaws in AM specimens.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗