Proposals of Guidelines for High Temperature Structural Design of Fast Reactor Vessels

Kasahara, Naoto; Satoh, Kenichiro; Tsukimori, Kazuyuki; Kawasaki, Nobuchika

doi:10.1115/pvp2010-25414

Cited by 5 publications

(2 citation statements)

References 2 publications

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“…Plastic deformation during cyclic fatigue can produce a dramatic reduction in creep lifetime. In general, either time fraction (stress-based) or ductility exhaustion (strain-based) approaches have been used to estimate creep-fatigue failure criteria [338][339][340]. It has been noted that the traditional ASME boiler and pressure vessel section III, Division 5 creep-fatigue methodology can produce very conservative predicted lifetimes in FM steels such as modified 9Cr-1Mo steel [341,342].…”

Section: Irradiation Creep and Creep-fatigue Interactionmentioning

confidence: 99%

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

et al. 2022

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Reduced activation ferritic-martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. Performance of these steels will deteriorate in-service due to neutron damage and transmutation-induced gases like helium/hydrogen at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the major irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (i) low temperature hardening-embrittlement (LTHE) - including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile to brittle transition temperature and fracture toughness, (ii) inter-mediate temperature cavity swelling, (iii) the effect of helium on LTHE and cavity swelling, (iv) irradiation creep and (v) tritium management issues are reviewed. The potential causes of LTHE are discussed that highlights need for advanced characterization techniques. Mechanical properties including tensile/Charpy impact of RAFM and ODS steels are compared to show that current generation of ODS steels also suffer from LTHE, and they can show irradiation hardening up to high temperatures, ~400-500 °C. To minimize this, future ODS steel development for FW/B specific application should target materials with lower Cr concentration (to minimize α’) and minimize other elements that could form embrittling phases under irradiation. RAFM steel designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand synergistic effects of helium on thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex including stresses due to magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multi-scale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility

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Section: Irradiation Creep and Creep-fatigue Interactionmentioning

confidence: 99%

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

et al. 2022

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“…As noted elsewhere [92][93][94][95], improved science-based high temperature design rules for irradiated structures are needed to replace current empirical and incomplete design criteria. These improved design rules would be of particular importance to take full advantage of the higher temperature capability of the new RAFM steels.…”

Section: Technical Challenges For Next-generation 8-9%cr Rafmsmentioning

confidence: 99%

Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications

et al. 2017

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Reduced activation ferritic/martensitic steels are currently the most technologically mature option for the structural material of proposed fusion energy reactors. Advanced nextgeneration higher performance steels offer the opportunity for improvements in fusion reactor operational lifetime and reliability, superior neutron radiation damage resistance, higher thermodynamic efficiency, and reduced construction costs. The two main strategies for developing improved steels for fusion energy applications are based on (1) an evolutionary pathway using computational thermodynamics modelling and modified thermomechanical treatments (TMT) to produce higher performance reduced activation ferritic/martensitic (RAFM) steels and (2) a higher risk, potentially higher payoff approach based on powder metallurgy techniques to produce very high strength oxide dispersion strengthened (ODS) steels capable of operation to very high temperatures and with potentially very high resistance to fusion neutron-induced property degradation. The current development status of these nextgeneration high performance steels is summarized, and research and development challenges for the successful development of these materials are outlined. Material properties including temperature-dependent uniaxial yield strengths, tensile elongations, high-temperature thermal creep, Charpy impact ductile to brittle transient temperature (DBTT) and fracture toughness behaviour, and neutron irradiation-induced low-temperature hardening and embrittlement and intermediate-temperature volumetric void swelling (including effects associated with fusionrelevant helium and hydrogen generation) are described for research heats of the new steels.

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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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Proposals of Guidelines for High Temperature Structural Design of Fast Reactor Vessels

Cited by 5 publications

References 2 publications

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications

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

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