High Resolution Transmission Electron Microscopy of Irradiation Damage in Inconel X-750

Judge, C. D.; Rajakumar, Heygaan; Korinek, Andreas; Botton, Gianluigi A.; Cole, James I.; Madden, James W.; Jackson, John H.; Freyer, Paula D.; Giannuzzi, L. A.; Griffiths, M.

doi:10.1007/978-3-319-67244-1_47

Cited by 7 publications

(24 citation statements)

References 23 publications

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“…Rate theory calculations [ 49 ] show that the net flux of point defects to boundaries (He atoms and vacancies) is dependent on the grain interior sink strength, as shown in Figure 13 and Figure 14 . In principle smaller cavities on the boundary will lead to a higher area coverage for a given stabilised cavity volume.…”

Section: Inter-granular Fracturementioning

confidence: 99%

“…The boundaries can receive a higher flux of point defects at higher temperatures if the sink density within the grain interior is lower at high temperatures than for lower temperatures. Modelling shows that to achieve a higher area coverage at higher temperatures there has to be either: (i) a lower sink strength in the grain interior; or (ii) He trapping at freely-migrating vacancies that have a higher steady-state concentration at lower temperatures [ 49 ]. Either or both of these factors will affect and lower the He and net vacancy flux to grain boundaries at the lower temperatures.…”

Section: Inter-granular Fracturementioning

confidence: 99%

“…Either or both of these factors will affect and lower the He and net vacancy flux to grain boundaries at the lower temperatures. The important result from rate theory is the trend of grain boundary area coverage with increasing dose and temperature showing a decreasing rate of coverage as a function of dose at high doses and an increase in coverage as a function of temperature [ 49 ]. The limiting factor in the evolution of the grain boundary cavity structure are the point defect fluxes from the grain interior, which are non-linear with dose.…”

Section: Inter-granular Fracturementioning

confidence: 99%

“…The model developed to predict the grain boundary area coverage by He-stabilised cavities for Inconel X-750 irradiated in a CANDU reactor [ 49 ], has been applied to the swelling and grain boundary cavity coverage for the PWR flux thimble studied by Fukuya et al [ 39 ], and Edwards et al [ 40 ], shown in Figure 10 . The cavity number density in the matrix for the flux thimble irradiated to 70 dpa at 315 °C is approximately 2 × 10 23 m −3 [ 40 ].…”

Section: Inter-granular Fracturementioning

confidence: 99%

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Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Griffiths

2021

Materials

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Austenitic stainless steels are used for core internal structures in sodium-cooled fast reactors (SFRs) and light-water reactors (LWRs) because of their high strength and retained toughness after irradiation (up to 80 dpa in LWRs), unlike ferritic steels that are embrittled at low doses (<1 dpa). For fast reactors, operating temperatures vary from 400 to 550 °C for the internal structures and up to 650 °C for the fuel cladding. The internal structures of the LWRs operate at temperatures between approximately 270 and 320 °C although some parts can be hotter (more than 400 °C) because of localised nuclear heating. The ongoing operability relies on being able to understand and predict how the mechanical properties and dimensional stability change over extended periods of operation. Test reactor irradiations and power reactor operating experience over more than 50 years has resulted in the accumulation of a large amount of data from which one can assess the effects of irradiation on the properties of austenitic stainless steels. The effect of irradiation on the intrinsic mechanical properties (strength, ductility, toughness, etc.) and dimensional stability derived from in- and out-reactor (post-irradiation) measurements and tests will be described and discussed. The main observations will be assessed using radiation damage and gas production models. Rate theory models will be used to show how the microstructural changes during irradiation affect mechanical properties and dimensional stability.

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Section: Inter-granular Fracturementioning

confidence: 99%

Section: Inter-granular Fracturementioning

confidence: 99%

Section: Inter-granular Fracturementioning

confidence: 99%

Section: Inter-granular Fracturementioning

confidence: 99%

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Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Griffiths

2021

Materials

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“…The stability of γ' precipitates play an important role in the hardening of alloy X-750 and are known to disorder and dissolve as a function of irradiation damage, thereby resulting in softening of the material [17]. The stability of γ' is therefore characterized as a function of irradiation dose to ascertain the degree of disordering and dissolution which has taken place.…”

Section: Stability Of γ' Precipitates In Irradiated Alloy X-750mentioning

confidence: 99%

Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)

Jackson

Heighes

Andresen³

2020

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The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10 19 n/cm 2 , 2.0 x 10 20 n/cm 2 , and 1.0 x 10 21 n/cm 2. These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10 19 n/cm 2 , comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60-80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project. Following completion of the irradiations at ATR in the water-cooled center flux trap, irradiation assisted stress corrosion cracking (IASCC), fracture toughness (FT), tensile testing, and transmission electron microscopy (TEM) studies were conducted at INL. Testing under normal water chemistry (NWC) conditions indicated a negligible effect on crack growth rate (CGR) at the medium and high fluence irradiation levels compared to the CGRs measured in unirradiated material. Under hydrogen water chemistry (HWC) conditions, only a modest increase in CGR was measured at the high fluence level compared to the medium fluence level. Tensile and FT measurements were far more sensitive to irradiation level, especially in the case of alloy X-750, with an increase of 21, 29, and 57% of yield strength compared to unirradiated levels for the low, medium, and high fluences, respectively, and an approximate 4, 18, and 41% reduction in FT for the low, medium, and high fluences, respectively. Additionally, TEM analyses showed an increase in dislocation loop size as a function of irradiation dose for both alloys, but a negligible change in loop density.

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Radiation Effects in Nickel-Based Alloys

Griffiths¹,

Boothby²

2020

Comprehensive Nuclear Materials

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High Resolution Transmission Electron Microscopy of Irradiation Damage in Inconel X-750

Cited by 7 publications

References 23 publications

Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)

Radiation Effects in Nickel-Based Alloys

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