800H Weldments and Crack Growth Rates of CGR Alloys
Update on FY24 crack growth rate and 800H weldment testing.
36 MATERIALS SCIENCE↗
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Temperature-dependent mechanism on hardening mitigation in thermomechanically processed 800H alloys under neutron irradiation
Irradiation induced hardening is commonly observed in structural materials, and is often associated with the degraded ductility. This study explores the role of thermomechanical processing (TMP) in mitigating irradiation-induced hardening in Incoloy 800H. The radiation response of solution-annealed and TMP-treated samples was systematically investigated to elucidate the underlying mechanisms of hardening reduction. Neutron irradiation was performed on solution-annealed and TMP-treated samples at 359, 431, and 580 °C, revealing consistently lower hardening in TMP samples. Microstructural characterization compared dislocation loops, cavities, clusters, and precipitates to assess their contributions to hardening. The strengthening contributions from individual microstructural features were estimated, and the microstructure-derived yield strength is consistent with the hardness-converted yield strength. The microstructure–properties correlation revealed the temperature-dependent primary hardening mechanism: dislocation loops dominated at lower temperature, whereas Ni3(Al,Ti) γ′ precipitates were predominant at higher temperature. TMP reduced radiation-induced hardening through two mechanisms: (1) increased sink strength from dislocations and Ti(C,N) precipitates limited dislocation loop formation at lower temperatures, and (2) reduced Ti solute within the matrix in TMP samples suppressed γ′ precipitate formation at higher temperatures. These findings identify the mechanisms governing radiation-induced hardening across different temperatures, which lays the foundation for TMP optimization for the development of advanced radiation-tolerant materials.
Zhong, Weicheng [ORNL] (ORCID:0000000231589271)↗
An elastoplastic finite element study of cylindrical plane-strain contact for steel/steel and Inconel 617/Incoloy 800H undergoing unidirectional sliding
The current work employs a two-dimensional plane strain finite element analysis to investigate the unidirectional sliding contact between a deformable half-cylinder and a deformable flat block. The unidirectional sliding is governed by a displacement-controlled action where the materials of the two contacting bodies are first set to identical steels at 20 ℃ and then to Inconel 617 and Incoloy 800H at 800 ℃. First, a normal interference (indentation) is applied, which is followed by unidirectional sliding. The von Mises stress distribution, plastic strain distribution, junction growth, normal force, tangential force, effective coefficient of friction, and scars on the surface of the block are obtained during the sliding motion. The leading edge of the contacting area and the bulk material under the leading edge experience large von Mises stresses. The large plastic strain is found on the surface of the block, and forms a “pocket” shape under the surface. The junction growth is also investigated, showing the direction of the growth is in the same direction of the tangential force that the weaker material experiences. The forces and the effective coefficient of friction are found to stabilize after a certain sliding distance, and the effective coefficient of friction converges to the coefficient of friction used in the model. Pileup is found on the surface of the block after a sufficient unidirectional sliding distance.
Yang, Huaidong↗
Initial development of viscoplastic constitutive model of Alloy 800H in support of the use of inelastic analysis methods for ASME Section III, Division 5, Class A applications
This report describes the development of a preliminary inelastic constitutive model for the thermomechanical behavior of Alloy 800H. The objective is to develop a model suitable for incorporating into Nonmandatory Appendix Z to Section III, Division 5 of the ASME Boiler & Pressure Vessel Code, which provides guidance and reference constitutive models for the ASME design by inelastic analysis rules for Class A components. The report describes the process of collecting experimental data, developing a mathematical form for the model, and training the model against the test data. The initial version of the model captures most of the relevant material deformation mechanisms, including dynamic strain aging effects. However, further development of the model form will be required to develop a final model suitable for ASME use.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
On massive carbide precipitation during high temperature low cycle fatigue in alloy 800H
The effect of strain rate on massive precipitation and the mechanism for the occurrence of massive precipitation of M23C6 in alloy 800H is investigated during elevated temperature low cycle fatigue testing. It was observed that large M23C6 platelets were in the vicinity of grain and incoherent twin boundaries. The strain controlled fatigue testing at higher strain rates that promoted cyclic hardening enabled massive precipitation to occur more easily.
CARBON SUPERSATURATION↗
Temperature and Strain-Rate Effects on Low-Cycle Fatigue Behavior of Alloy 800H
The effects of strain rate (4 x 10(exp -6) to 4 x 10(exp -3)/s) and temperature on the Low-Cycle Fatigue (LCF) behavior of alloy 800H have been evaluated in the range 750 C to 950 C. Total axial strain controlled LCF tests were conducted in air at a strain amplitude of +/- 0.30 pct. LCF life decreased with decreasing strain rate and increasing temperature. The cyclic stress response behavior showed a marked variation with temperature and strain rate. The time- and temperature- dependent processes which influence the cyclic stress response and life have been identified and their relative importance assessed. Dynamic strain aging, time-dependent deformation, precipitation of parallel platelets of M(23)C6 on grain boundaries and incoherent ledges of twins, and oxidation were found to operate depending on the test conditions. The largest effect on life was shown by oxidation processes.
Rao, K. Bhanu Sankara↗
Mechanisms of High-Temperature Fatigue Failure in Alloy 800H
The damage mechanisms influencing the axial strain-controlled Low-Cycle Fatigue (LCF) behavior of alloy 800H at 850 C have been evaluated under conditions of equal tension/compression ramp rates (Fast-Fast (F-F): 4 X 10(sup -3)/s and Slow-Slow (S-S): 4 X 10(sup -5)/s) and asymmetrical ramp rates (Fast-Slow (F-S): 4 x 10(sup -3)/s / 4 X 10(sup -5/s and Slow-Fast (S-F): 4 X 10(sup -5) / 4 X 10(sup -3)/s) in tension and compression. The fatigue life, cyclic stress response, and fracture modes were significantly influenced by the waveform shape. The fatigue lives displayed by different loading conditions were in the following order: F-F greater than S-S greater than F-S greater than S-F. The fracture mode was dictated by the ramp rate adopted in the tensile direction. The fast ramp rate in the tensile direction led to the occurrence of transgranular crack initiation and propagation, whereas the slow ramp rate caused intergranular initiation and propagation. The time-dependent processes and their synergistic interactions, which were at the basis of observed changes in cyclic stress response and fatigue life, were identified. Oxidation, creep damage, dynamic strain aging, massive carbide precipitation, time-dependent creep deformation, and deformation ratcheting were among the several factors influencing cyclic life. Irrespective of the loading condition, the largest effect on life was exerted by oxidation processes. Deformation ratcheting had its greatest influence on life under asymmetrical loading conditions. Creep damage accumulated the greatest amount during the slow tensile ramp under S-F conditions.
BhanuSankaraRao, K.↗
Microstructural effect on neutron irradiation response of alloy 800H
No abstract prepared.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
An Initial Assessment of the Creep-Rupture Strengths for Weldments with Alloy 800H Base Metal and Alloy 617 Filler Metal
The expected minimum stress-to-rupture of the weld is a function of the stress rupture factor (R) and the expected minimum stress-to-rupture (Sr) of the base metal.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Effects of heat treatment on corrosion fatigue and stress corrosion crack growth of additive-manufactured Alloy 800H in high-temperature water
Not Available
Yang, Jingfan (ORCID:0000000254407872)↗
Characterization of in-situ ion irradiated Fe-21Cr-32Ni austenitic model alloy and alloy 800H at low
Explore the source record for details and available documents.
Ayanoglu, M.↗
Status of Alloy 800H in Considerations for the Gen IV Nuclear Energy Systems
Alloy 800 H is currently under consideration for applications in the next generation nuclear plant (NGNP) at operational temperatures above 750°C. This paper first describes service requirements of the nuclear system for structural materials; and then an extensive review of Alloy 800 H is given on its codification with respect to development and research history, mechanical behavior and design allowables, metallurgical aging resistance, environmental effect considerations, data requirements and availability, weldments, as well as many other aspects relevant to the intended nuclear application. Finally, further research and development activities to support the materials qualification are suggested.
36 MATERIALS SCIENCE↗
Strain Rate Sensitivity of Alloys 800H and 617
The flow stress of many materials is a function of
36 MATERIALS SCIENCE↗
Novel Dissimilar Joints Between Alloy 800H and 2.25%Cr and 1%Mo Steel
Dissimilar metal joints between ferritic steels and nickel base alloys are currently fabricated using conventional arc welding processes with selected filler metal combinations. The dissimilar metal joints contain abrupt changes in composition over a relatively small distance. Many years of service at elevated temperatures has shown that these welds are susceptible to creep and creep fatigue failures. The primary mechanism for these creep failures involves carbon diffusion from the ferritic steel to the nickel base alloy. As a result, a carbon depleted zone is created that contains very few stable carbides. This work used additive manufacturing (AM) technologies as a highly controllable means for reducing carbon migration through theoretically designed, compositionally graded transition joints guided by appropriate thermodynamic, kinetic and heat transfer and fluid flow modeling. The contents of the report include the objectives and tasks set forth by the PI and collaborators, major technical accomplishments throughout the research and additional details in the form of technical publications resulting from the NEUP grant and reports from the collaborating university and national laboratory.
36 MATERIALS SCIENCE↗
Report on FY 2021 testing in support of integrating Alloy 800H and 2.25Cr-1Mo for EPP analysis
The use of simplified Elastic-Perfectly Plastic (EPP) design methods avoids the complexities and limitations of the elastic analysis approach in evaluating strain limits and creep-fatigue damage for Class A metallic coolant boundary components in ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B. No stress classification and linearization are required in the EPP methods. Further, the strain limits and creep-fatigue damage evaluation procedures based on elastic analysis results were developed for the reactor operating conditions where uncoupled, rate-independent plasticity and stationary creep can be used to approximate the deformation in the component adequately. But at very high temperatures, the deformation is rate-dependent, and it can only be approximated adequately by unified viscoplastic model. Thus, for example, the elastic analysis approach cannot be used to evaluate strain limits and creep-fatigue damage for Alloy 617 above 650 °C. The EPP methods remove the temperature restriction. They can be used to evaluate strain limits and creep-fatigue damage for the full range of temperatures.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Experimental Basis for the Extension of Elastic-Perfectly Plastic Strain Limits Evaluation Procedure of ASME Section III, Division 5 Code Case N-861 to Grade 91, Alloy 800H and 2.25Cr-1Mo
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Wang, Yanli↗
Second Annual Progress Report on Correlation Between Microstructure and Mechanical Properties of Neutron-Irradiated Ferritic-Martensitic and Austenitic Steels
Ferritic-martensitic steels G92-2b (an optimized Grade 92 heat), NF616 and T91, and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) and the Advanced Test Reactor (ATR) of Idaho National Laboratory (INL). Selected G92-2b samples were irradiated up to 14.66 dpa in the HFIR at two temperature ranges: 400–496.7°C and 683.3– 720°C. NF616 and T91 were irradiated in the ATR up to 8.16 dpa with the irradiation temperatures ranged from 241°C to 447.5°C. Alloy 800H and 800H-TMP samples were irradiated in both the HFIR and the ATR. Selected 800H and 800H-TMP samples had HFIR irradiation to 1.28 dpa at 580°C and ATR irradiation up to 9.12 dpa at 359°C to 431°C. Vickers hardness measurements, fractography, and microstructural characterization were performed on the selected samples in the Low Activation Materials Design and Analysis (LAMDA) laboratory. Radiation-hardening of G92-2b was observed at the lower doses and lower irradiation temperatures (400- 496.7°C), with GB03 (0.52 dpa at 400°C) and GB04 (7.44 dpa at ~490°C) showing ~12% and ~8% hardening, respectively. Softening by ~14% was observed for GB05 (14.66 dpa at 496.7°C). Radiationsoftening of G92-2b was more prevalent at the higher irradiation temperatures (683.3-~720°C), with GB10 (0.46 dpa at 683.3°C), GB11 (7.44 dpa at ~720°C), and GB12 (14.63 dpa at ~720°C) showing ~8%, ~8%, and ~40% softening, respectively. Radiation-hardening of NF616 and T91 was observed with the hardness increased by ~37% to ~65% depending on the irradiation doses and irradiation temperatures. Within the studied irradiation conditions of NF616 and T91, samples with a higher dose had a larger hardness after irradiation. All the tested alloy 800H and 800H-TMP samples in this work showed radiation-hardening by ~96±7% to ~152±10%. Alloy 800H-TMP tended to have slightly smaller radiation-hardening than alloy 800H. The fractography results of G92-2b sample GB03, GB10, and GB11, together with the previously characterized fractography of GB04, GB05, and GB12, indicated that the ductility of G92-2b was maintained up to 14.66 dpa at the lower irradiation temperatures of 400-496.7°C, while some loss of ductility (less necking) was observed for higher doses at the higher irradiation temperatures of 683.3-720°C. This agrees with the previously reported tensile test results of G92-2b, where the elongation of G92-2b was reduced at higher doses at the higher irradiation temperatures. Dimple sizes increased at higher doses, which are more evident at the higher irradiation temperatures of 683.3-~720°C. The fractography of NF616 sample D2 (2.96 dpa at 291.5°C), D4 (5.91 dpa at 359°C), and D6 (8.16 dpa at 431°C) indicated loss of ductility with negligible necking for sample D2, while ductile failure for samples D4 and D6. Fractography of alloy 800H and 800H-TMP samples in various irradiation conditions showed ductile failure with obvious necking. Dimples were observed, with some of them containing large Ti-rich particles, in all the 800H/800H-TMP samples. Electron backscatter diffraction characterization of GB12 indicated the recovery of the lath structure, which was generally replaced by an equiaxed grain structure. Transmission electron microscopy (TEM) characterization showed the presence of frequent M 23 C 6 (M = primarily Cr), MX (M = primarily V), spherical Nb(C,N) precipitates, and occasional Laves phase precipitates in the G92-2b samples. MX precipitates with sizes of 20-30 nm were observed at boundaries of smaller grains, indicating the pinning effect of the V-rich precipitates. The lath structure recovery was more evident at the higher irradiation temperatures (683.3-~720°C), with decreased densities of line dislocations and M23C6 precipitates. The irradiated T91 (TA04) showed the growth of M23C6 precipitates to 101 ± 40 nm from the initial 68 ± 22 nm in the unirradiated condition. Dislocation loops of both {100} and {111} types were present in TA04. TEM characterization was also performed on the irradiated 800H (N4, N5, N6, and AR2) and 800H-TMP (P4, P5, P6, and HG1). Accumulation of M 23 C 6 precipitates at grain boundaries was observed in all the xii 800H/800H-TMP samples, and the presence of Ti(C,N) precipitates at grain boundaries and in the matrix was observed in the irradiated 800H-TMP. Some Ti(C,N) precipitates are embedded in the M 23 C 6 precipitates, maintaining specific orientation relationships between the precipitates and between the precipitate and matrix. In addition, nanoscale Si-rich clusters were observed in the matrix of all the 800H/800H-TMP samples, with EDS Si maps tending to have a lower contrast in 800H-TMP samples. Atom probe tomography was conducted on the same samples, supported by a Rapid Turnaround Examination project under Nuclear Science User Facilities. The results are being analyzed to be integrated with the TEM results for a confident description of the γ’ precipitates. Dislocation loops also formed in all the 800H/800H-TMP samples. The density and the average size of dislocation loops were quantified to be in the order of 10 22 – 10 23 m -3 and 11.7 – 15.9 nm, respectively, in the ATR-irradiated 800H/800H-TMP samples. The loop densities in irradiated 800H were higher than that in irradiated 800H-TMP under the same irradiation conditions. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.
36 MATERIALS SCIENCE↗
Report on partially complete post-irradiation-examination of the INL samples
Ferritic-martensitic steels (NF616 and T91), and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and their microstructures and mechanical properties were characterized. Ferritic-martensitic steels were irradiated up to 8.16 dpa in the irradiation temperatures ranging from 241 °C to 447.5 °C. Selected 800H and 800H-TMP samples were also irradiated up to 9.12 dpa at 359°C to 431°C. Vickers hardness measurement, fractography, and microstructural characterization were performed on the selected samples in the LAMDA lab. Radiation-induced hardening of NF616 and T91 was observed in the hardness measurements which show ~37% to ~65% increases, depending on irradiation dose and irradiation temperature. Within the irradiation conditions studied in this work, samples with a higher dose had a higher hardness. All the tested alloy 800H and 800H-TMP in this work showed irradiation-hardening by ~96±7% to ~152±10%. Alloy 800HTMP tended to have slightly smaller irradiation-hardening than alloy 800H. The fractography of NF616 samples D2 (2.96 dpa at 291.5°C), D4 (5.91 dpa at 359°C), and D6 (8.16 dpa at 431°C) indicated loss of ductility with negligible necking for the sample D2, yet obvious ductile failure for samples D4 and D6. Fractography of alloy 800H samples N4 (7.27 dpa at 359°C), N5 (3.9 dpa at 451.5°C), and N6 (9.01 dpa at 431°C), and alloy 800H-TMP samples P4 (7.36 dpa at 359°C), P5 (3.95 dpa at 451.5°C), and P6 (9.12 dpa at 431°C) indicated that all samples failed in a ductile mode, with obvious necking. Dimples were observed where large Ti-containing particles were observed for the samples. TEM characterization was performed on alloy 800H (N4, N5, N6) and 800H-TMP (P4, P5 and P6). Accumulation of M23C6 precipitates at grain boundaries was observed in all samples, and the presence of Ti(C,N) precipitates at grain boundaries and in the matrix was observed in alloy 800H-TMP. M23C6 precipitates were found to have the cube-on-cube orientation relationship with the matrix. In addition, nanoscale Si-rich clusters were observed in the matrix of all samples. Atom probe tomography was conducted on the same samples, and their results are being analyzed to be integrated with the TEM results for a confident description of the clustering behavior. Due to the low irradiation temperature (up to 451.5°C), dislocation loops formed in all samples, with the density and the average size of dislocation loops in the order of 1022 – 1023 m-3 and 11.7 – 15.9 nm, respectively. The loop densities in alloy 800H were higher than that in alloy 800H-TMP under the same irradiation conditions. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗