Overview of Indian activities on fusion reactor materials

Banerjee, S.

doi:10.1016/j.jnucmat.2014.06.009

Cited by 23 publications

(6 citation statements)

References 52 publications

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“…Indeed RAFM steels developed by some of the 'newer' members of the worldwide fusion research community have moved the basic EUROFER/F82H type of RAFM steels towards a 2% tungsten content, with some demonstrated improvement in creep and tensile strength properties [58].…”

Section: High Temperature Ferritic-martensitic Steelsmentioning

confidence: 99%

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

Stork

Agostini

Boutard

et al. 2014

Journal of Nuclear Materials

221

135

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The findings of the EU 'Materials Assessment Group' (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision.A DEMO phase I with a 'Starter Blanket' and 'Starter Divertor' is foreseen: the blanket being capable of withstanding ≥2MW.yr.m -2 fusion neutron fluence (~20 dpa in the frontwall steel). A second phase ensues for DEMO with ≥5MW.yr.m -2 first wall neutron fluence.Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction § Corresponding author. Address as 1. email: derek.stork@btinternet.com *Manuscript Click here to view linked References with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials.The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650ºC for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (~290-320ºC), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R&D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (~200-350ºC) that could be realised, as a baseline-concept, using tungsten on a copper-alloy substructure. The difficulty of establishing design codes for brittle tungsten puts great urgency on the development of a range of advanced ductile or strengthened tungsten and copper compounds.Lessons learned from Fission reactor material development have been included, especially in safety and licensing, fabrication/joining techniques and designing for in-vessel inspection. The technical basis of using the ITER licensing experience to refine the issues in nuclear testing of materials is discussed.Testing with 14MeV neutrons is essential to Fusion Materials development, and the Roadmap requires acquisition of ≥30 dpa (steels) 14MeV test data by 2026. The value and limits of pre-screening testing with fission neutrons on isotopically-or chemically-doped steels and with ion-beams are evaluated to help determine the minimum14 MeV testing programme requirements.

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Section: High Temperature Ferritic-martensitic Steelsmentioning

confidence: 99%

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

Stork

Agostini

Boutard

et al. 2014

Journal of Nuclear Materials

221

135

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“…The compatibility of the construction material's properties and the working conditions of the nuclear reactors are still major concerns for the long-term safety and environmental degradation [3]. Currently, material scientists are facing major challenges imposed by the increasing need for novel materials that can withstand the extreme working conditions of the future generations of nuclear reactors [4], such as fusion and fission nuclear power [5], and this is a priority because any material failure can result in a severe accident [6]. Viswanathan et al reported the challenges and advances in nanocomposite processing techniques, and that the manufacturing process of the materials has a direct impact on their final properties.…”

Section: Introductionmentioning

confidence: 99%

Novel Alumina Dispersion-Strengthened 316L Steel Produced by Attrition Milling and Spark Plasma Sintering

et al. 2023

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Alumina dispersion-strengthened 316L stainless steels were successfully produced using attrition milling and spark plasma sintering. Two different composites (316L/0.33 wt% and 316L/1 wt% Al2O3) were prepared by powder technology. The attrition milling produced a significant morphological transformation of the globular 316L starting powders and provided a homogeneous distribution of the nanosized alumina particles. The XRD results confirmed that the 316L steel was an austenitic γ-Fe3Ni2. The formation of a ferrite α-Fe phase was detected after milling; this was transformed to the austenitic γ-Fe3Ni2 after the sintering process. The addition of nanosized alumina particles increased the composites’ microhardness significantly to 2.25 GPa HV. With higher amounts of alumina, the nanosized particles tended to agglomerate during the milling process. The friction coefficient (FC) of the 316L/0.33 wt% Al2O3 and the 316L/1 wt% Al2O3 decreased because of the increase in the composite’s hardness; FC values of 0.96, 0.93 and 0.85, respectively, were measured respectively for the 316L reference, the 316L/0.33 wt% and the 316L/1 wt% Al2O3. The 316L/0.33 wt% Al2O3 composite had a higher flexural strength of 630.4 MPa compared with the 316L/1 wt% Al2O3 with 386.6 MPa; the lower value of the latter was related the agglomeration of the alumina powder during attrition milling.

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“…The Indian lead lithium ceramic breeder test blanket module (LLCB-TBM) concept uses Pb-Li as the coolant, neutron multiplier and tritium breeder, which will be tested for its performance at the upcoming fusion reactor, ITER [2,3]. However, Indian reduced activation ferritic martensitic steel (IN RAFMS) is the candidate structural material for this TBM [2,4]. IN RAFMS is an altered version of the modified 9Cr-1Mo steel (P91) where elements producing long half-life radioisotopes, like molybdenum and niobium, have been substituted by safer alternatives, such as tungsten and tantalum, respectively [5].…”

Section: Introductionmentioning

confidence: 99%

Effect of magnetic field on corrosion behavior of Indian reduced activation ferritic/martensitic steel in liquid Pb–Li

et al. 2019

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In the present study, the effect of a magnetic field on the corrosion of Indian reduced-activation ferritic/martensitic steel (IN RAFMS) in flowing lead–lithium eutectic (Pb–Li) has been studied in an electromagnetic-pump-driven loop (EMPPIL-M). The loop was operated for 2800 h at an average flow velocity of 1.5 m s−1 with a 0.5 T magnetic field over the IN RAFMS samples. The corrosion rate in the presence of a magnetic field at a temperature of 773 K was found to be 1.3 times higher than that observed in the absence of a magnetic field. Non-uniform corrosion, resulting in the formation of distinct surface features, was observed over the surface of the sample located inside the magnetic field. Pb–Li attack in the presence of a magnetic field was not only confined to the prior austenite and martensitic lath boundaries (as in the absence of magnetic field), but also occurred in the intra-lath regions resulting in the formation of substructures. The development of electric potential over metallic samples and significant gradients in Pb–Li flow due to the magnetohydrodynamic effect played a major role in the formation of surface features and accelerating the corrosion rate within the magnetic field. On the other hand, the increase in the depletion of iron from the steel matrix observed in the presence of the magnetic field was a result of its direct interaction with the field parameters.

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Overview of Indian activities on fusion reactor materials

Cited by 23 publications

References 52 publications

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

Novel Alumina Dispersion-Strengthened 316L Steel Produced by Attrition Milling and Spark Plasma Sintering

Effect of magnetic field on corrosion behavior of Indian reduced activation ferritic/martensitic steel in liquid Pb–Li

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