On the effects of irradiation and helium on the yield stress changes and hardening and non-hardening embrittlement of ∼8Cr tempered martensitic steels: Compilation and analysis of existing data

Yamamoto, Takuya; Odette, G.R.; Kishimoto, Hirotatsu; Rensman, J.W.; Miao, Pifeng

doi:10.1016/j.jnucmat.2006.05.041

Cited by 140 publications

(94 citation statements)

References 80 publications

(144 reference statements)

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“…In the irradiation hardening dominated regime below about 400°C, and in the absence of large quantities of helium, the coarse scale microstructures and the corresponding "*(T) and V*, are not effected by irradiation [17][18][21][22]. Thus !T o can be estimated from tensile, or even indentation hardness, tests.…”

Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

“…We have derived !" y models based on fits of physically motivated equations for h(T i ,dpa) fit to a large database on 9Cr tempered martensitic steels (TMS) irradiated at a wide range of temperatures and dose for both individual alloys and overall genetic alloy class trends [21][22]. Thus Equation 1 provides a good basis to estimate !T o for a similar range of alloys and irradiation condition s in the hardening dominated embrittlement regime.…”

Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

“…Thus Equation 1 provides a good basis to estimate !T o for a similar range of alloys and irradiation condition s in the hardening dominated embrittlement regime. Finally, Equation 1 can be used to estimate the !T o associated with different loading rates based on a strain rate compensated temperature shift [21][22]. Figure 3 shows the predicted dpa dose dependence of !T o assuming C o = 0.6C/MPa, at various irradiation temperatures between 60 and 400°C.…”

Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

“…An example, of model guided alloy design and optimization is given in the next section. Finally, the hardening dominated !T o MC model provides a basis to assess embrittlement (!T o ) that are not due to hardening alone, or even occur in the absence of hardening [18,[21][22]. Such non-hardening embrittlement mechanisms, including very large !T o caused by a combination of large !"…”

Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

“…Data on irradiation hardening and embrittlement of 8-10Cr normalized and tempered martensitic steels (TMS) alloys has been compiled from the literature [18,[21][22]. Most of the embrittlement data, covering a wide range of irradiation conditions, is for sub-sized Charpy tests.…”

Section: Non-hardening Irradiation Assisted Thermal and Helium Embritmentioning

confidence: 99%

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Odette¹,

Lucas²

2005

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Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

Section: An Embrittlement !T O Prediction Model For the Irradiation Hmentioning

confidence: 99%

Section: Non-hardening Irradiation Assisted Thermal and Helium Embritmentioning

confidence: 99%

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In-Service Design & Performance Prediction of Advanced Fusion Material Systems by Computational Modeling and Simulation

Odette¹,

Lucas²

2005

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Plasma‐material Interactions Problems and Dust Creation and Re‐suspension in Case of Accidents in Nuclear Fusion Plants: A New Challenge for Reactors like ITER and DEMO

Malizia

Poggi

Ciparisse

et al. 2016

Advanced Surface Engineering Materials

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Given the urgent need to converge on precise guidelines for accident management\ud in nuclear fusion plants, in this paper, the authors will analyze the problem\ud related to the choice of possible candidate materials for the nuclear fusion plants\ud like International Thermonuclear Experimental Reactor (ITER), DEMOnstration\ud power plant (DEMO), or PROTOtype power plant (PROTO). Fusion power is\ud a promising long-term candidate to supply the energy needs of humanity. From\ud the safety point of view, nuclear fusion holds inherent and potential safety advantages\ud over other energy sources. In magnetic confinement devices, the plasma\ud edge and surrounding material surfaces provide a buffer zone between the hightemperature\ud conditions in the plasma core and the normal “terrestrial” environment.\ud The interaction between the plasma edge and the surrounding surfaces\ud profoundly influences the conditions in the plasma core and is the principal key\ud engineering issue. Robust solutions to plasma–material interactions (PMIs) issues\ud are required to realize a commercially attractive fusion reactor. PMIs critically\ud affect tokamak operation in many ways. Erosion by the plasma determines the\ud lifetime of plasma-facing components (PFCs) and generates a source of impurities,which cool and dilute the plasma. Deposition of material onto PFCs alters their\ud surface composition and can lead to long-term accumulation of large in-vessel\ud tritium inventories. Retention and recycling of hydrogen from PFCs affect fueling\ud efficiency, plasma density control, and the density of neutral hydrogen in the\ud plasma boundary, which impacts particle and energy transport. Dusts are currently\ud produced in the existing plants like JET (and will also be in the future ones\ud like ITER, DEMO, and PROTO) by PMIs. Thus, the issues related to the PFCs of\ud the first wall of the nuclear fusion plants area topic that the Quantum Electronics\ud and Plasma Physics Research Group have been studying for more than a decade.\ud Regarding the selection of PFCs, there are two strong school of thoughts that\ud work on fusion plants materials for the first wall: (1) one composed by experts\ud working mainly on nuclear fusion and that are focusing their attention on certain\ud materials (like tungsten, beryllium, carbon, tritium, stainless steel, and other),\ud and (2) the other one composed of experts working mainly on the fission plants\ud that want to demonstrate the “dual-use” characteristics of the “fission materials”\ud for fusion plants applications. In this review paper, the authors will (1) review\ud both the approaches mentioned above comparing the main characteristics of all\ud the proposed and tested materials, (2) analyze the consequences (experimentally\ud revealed and numerically predicted) of PMIs that provokes serious damages to the\ud plants and dangerous consequences like radioactive/toxic dust production, and\ud (3) analyze (also thanks to our experimental activities on STARDUST-Upgrade\ud facility) the risks (for th...

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Microstructure of High‐Chromium Ferritic–Martensitic Steels for Next‐Generation Reactors

Košovský,

Dekan,

Sedlačková

et al. 2023

Physica Status Solidi (b)

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High‐chromium steels are key alloys used for the construction of technological devices in industry. Stainless steels are suitable for components that are exposed to a corrosive environment for a long time because chromium has anticorrosion properties due to segregation of chromium and the formation of a passivation layer. The physicochemical properties of the surface and the bulk of the material as well are determined by microstructure. Herein, steel NS219 is focused on, where the chromium concentration is around 13.5 wt%. In order to study the microstructure of steel, Mössbauer spectroscopy is used. Experimental results are evaluated using the binomial distribution model of the probability distribution of atoms in the nearest neighbor of the resonant atom 57Fe. Obtained spectral parameters, viz., the average magnetic hyperfine field, the average isomer shift, and the probability of the atomic configuration with no impurity atoms in the two‐shell vicinity of the iron atoms, reach saturation values from which the solubility limit of chromium in iron can be determined. On the other hand, the solubility limit of iron in Cr‐rich phase can be estimated from the value of the isomer shift of the single‐line in the spectrum annealed for the longest time.

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On the effects of irradiation and helium on the yield stress changes and hardening and non-hardening embrittlement of ∼8Cr tempered martensitic steels: Compilation and analysis of existing data

Cited by 140 publications

References 80 publications

In-Service Design & Performance Prediction of Advanced Fusion Material Systems by Computational Modeling and Simulation

In-Service Design & Performance Prediction of Advanced Fusion Material Systems by Computational Modeling and Simulation

Plasma‐material Interactions Problems and Dust Creation and Re‐suspension in Case of Accidents in Nuclear Fusion Plants: A New Challenge for Reactors like ITER and DEMO

Microstructure of High‐Chromium Ferritic–Martensitic Steels for Next‐Generation Reactors

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