Development of design Criteria for ITER In-vessel Components

Sannazzaro, G.; Barabash, V.; Kang, Suk-Jin; Fernández, Elena Cuasante; Kalinin, G.; Obushev, A.; Martínez, V.J.; Vázquez, Irma Isabel De León; Fernández, F.; Guirao, J.

doi:10.1016/j.fusengdes.2013.01.019

Cited by 32 publications

(8 citation statements)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The total elongations and fracture strains remain high (⩾10%) after irradiations around ∼2.5-3 dpa, and typically saturate between ∼5%-10% by ∼10-12 dpa (figure 6(b)). From a power-plant design point of view such as the descriptions in ITER structural design criteria for in-vessel components [61,155] or RCC-MRx, plastic flow localisation due to plastic instability in neutron irradiated RAFM steels is related to the loss of UE [156]-which presents the first failure mode, while ductility exhaustion induced local fracture is the second mode of component failure that is linked to the loss of total elongation [61]. The second mode of deformation is typically not a major issue for Gen-I RAFM steels because they retain sufficiently high total elongations (⩾5%, see figure 6(b)), and tensile failures are largely ductile.…”

Section: Low Temperature Hardening-embrittlement (Lthe) In Rafm and O...mentioning

confidence: 99%

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

et al. 2022

View full text Add to dashboard Cite

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

show abstract

Section: Low Temperature Hardening-embrittlement (Lthe) In Rafm and O...mentioning

confidence: 99%

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

et al. 2022

View full text Add to dashboard Cite

show abstract

“…This section will summarize the analysis approach and the obtained results. The structural design criteria, the material allowable and the verification methods used are taken from the SDC-IC Code [5] for assessing the CIT and the RCC-MR, Edition 2007 [6,7], including July 2017 Erratum for the bolts. The main loads acting on the component are (in no particular order): dead weight of the CIT, seismic acceleration forces due to a seismic level 2 earthquake, first plasma disruption generating halo and eddy currents, baking of the VV, plasma operation loads by direct radiative heating and ECRH stray radiation (max.…”

Section: System Load Specifications and Structural Integritymentioning

confidence: 99%

Final design of the ITER bolometer first plasma component: cable installation template

et al. 2021

View full text Add to dashboard Cite

The ITER bolometer cable installation template is a first plasma component which main function is to provide a temporary fixation of the mineral insulated cables and to protect them through first plasma operation inside the vacuum vessel until the final bolometer cameras are installed in the second assembly phase. In this paper, the final design of the cable installation template will be presented, the assembly procedures and important functional design details in order to be compliant with its interfaces, the vacuum vessel, the flux loops, the electrical services, the first plasma components and the assembly. Requirements that drive the design will be explained, along with the technical solutions chosen to fulfill them. Important sub-components will be presented in more detail, such as the boss fixation, which will compensate geometrical uncertainties of the as-built boss positioning, as well as the mineral insulated cable holder, which will prepare the electrical termination and facilitate the installation of final vacuum vessel cameras during the second assembly phase of ITER. Additionally, the step-by-step assembly plan agreed with the team of ITER machine assembly and tooling will be described. A custom developed boss welding tool will be described as well. The system load specifications applicable for this first plasma component will be presented as well and an overview of the structural integrity analysis report will be given. It will be shown that most loads on the component are very limited and there is no risk of structural failure or loss of relevant system functions. The paper concludes with a summary of the work and an overview of the procurement and delivery status to ITER.

show abstract

“…Table 2 illustrates the maximal deformation and m (P m is the general primary membrane equivalent stress, P L is the local primary membrane equivalent stress, P b is the primary bending equivalent stress). In a conservative way, the maximum P m is taken as the von Mises stress to calculate the margin [11,12]. However, the maximal deformation, 110.81 mm, is infeasible from the engineering design viewpoint.…”

Section: Design By Analysis Of the Manifold With Current Epsmentioning

confidence: 99%

Recent developments in the structural design and optimization of ITER neutral beam manifold

Cao¹,

Pan²,

Xia³

et al. 2018

Plasma Sci. Technol.

View full text Add to dashboard Cite

This paper describes a new design of the neutral beam manifold based on a more optimized support system. A proposed alternative scheme has presented to replace the former complex manifold supports and internal pipe supports in the final design phase. Both the structural reliability and feasibility were confirmed with detailed analyses. Comparative analyses between two typical types of manifold support scheme were performed. All relevant results of mechanical analyses for typical operation scenarios and fault conditions are presented. Future optimization activities are described, which will give useful information for a refined setting of components in the next phase.

show abstract

Development of design Criteria for ITER In-vessel Components

Cited by 32 publications

References 0 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

Final design of the ITER bolometer first plasma component: cable installation template

Recent developments in the structural design and optimization of ITER neutral beam manifold

Contact Info

Product

Resources

About