Comparative thermal cyclic testing and strength investigation of different Be/Cu joints

Gervash, A.; Giniyatulin, R.; Komarov, V.; Mazul, I.; Litunovsky, N.V.; Ganenko, A.A; Vainerman, A. A.; Fedotov, V. T.; Давыдов, Д. А.; Залавутдинов, Р. Х.

doi:10.1016/s0920-3796(98)00218-x

Cited by 15 publications

(5 citation statements)

References 5 publications

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“…7,[11][12][13][14] A new technology, fast brazing, has been developed by the Efremov institute (Russia). 15) Here the rapid temperature ramping rate (typically 160-180 K/min. in heating phase and 20 K/min.…”

Section: Be Materials and The Componentsmentioning

confidence: 99%

“…A number of electron beam facilities have been constructed, JUDITH 63) in Forschungszentrum Juelich (Germany), JEBIS, 64) OHBIS 62) in JAERI, ACT 65) in NIFS, EBHT 66) in Kyushu University (Japan), FE200 67) in CEA (France), TSEFFY 15) in Efremov Institute (Russia), EBTS, EB1200 21) in Sandia (USA), and an electron beam facility in SWIP (China). 68) In this paper, recent results of ITER relevant high heat flux testing performed by three of those electron beam facilities, JUDITH, JEBIS and OHBIS are mainly summarized.…”

Section: Static Heat Load Testing By Electron Beammentioning

confidence: 99%

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ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m 2 in the divertor targets and around 0.5 MW/m 2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

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Section: Be Materials and The Componentsmentioning

confidence: 99%

Section: Static Heat Load Testing By Electron Beammentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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“…Notably, the inner wall of the toroidal working chamber, a critical component, is constructed using beryllium [11][12][13][14][15][16][17][18][19][20]. Physicists involved in the development determined the optimal design for the blanket, a three-layer composite consisting of beryllium, a copper alloy, and stainless steel with an internal channel system for water cooling (Fig.…”

Section: Introduction Introductionmentioning

confidence: 99%

“…The areas where the constituents of the "beryllium -copper alloy -stainless steel" composite solidify should exhibit high thermal strength and thermal conductivity. Significant progress has been achieved in bonding beryllium and copper through brazing technology [18][19][20]. Nevertheless, the diffusion welding technique offers certain advantages [21][22][23][24] and, according to the authors, holds potential for further enhancing the performance properties of these composites.…”

Section: Introduction Introductionmentioning

confidence: 99%

Establishing theoretical foundations for predicting the structural and morphological characteristics of diffusion-welded joints of the beryllium–copper composite

Syrnev,

Maslennikov,

Semilutskaya

2024

Izv. VUZ. Poroshk. Met.

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The paper presents the results of theoretical and experimental studies regarding the quality of diffusion welding of the beryllium–copper composite. Numerical investigations of the parameters of heterodiffusion of diffusants and the thickness of the Be–Cu pair welded joint under varying temperature-time conditions were conducted. The analytical examinations revealed that the thickness of the diffusion weld at the Be–Cu joint varies between 26 and 345 µm, with the temperature increasing from 800 to 1000 °C and the holding time ranging from 20 to 120 min. The calculated layer thickness during the diffusion welding of a Be–Cu pair at 800 °C for 2 h is 65 µm, with 15 µm on the beryllium side and 50 µm on the copper side. Notably, a CuBe3 intermetallic compound zone can form in the diffusion weld, which should be considered an adverse factor that reduces the mechanical properties. To theoretically substantiate the modification of the structure and properties of the diffusion zone, a numerical study of welding was carried out using a 10 μm thick nickel foil spacer, which is readily soluble in beryllium. It was demonstrated that after exposure to temperature-time conditions at 900 °C for 20 min, a 50 µm wide diffusion-bonded joint is formed. Its structure includes two single-phase zones of solid solutions based on copper and beryllium, as well as two two-phase regions consisting of solid solutions hardened with intermetallic compounds. Since the weld lacks structural zones consisting solely of intermetallic compounds (unlike when welding the Be–Cu diffusion pair), there are grounds to anticipate a reduction in the embrittling effect on the weld. The results obtained from the analytical studies can serve as the foundation for a theoretical prediction method for assessing the quality of diffusion welding of the beryllium–copper composite.

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“…More than 10 different joining techniques were tested in Russia for Be-Cu joints in order to find an appropriate (non-silver) solution [5]. For ITER PFC application amorphous brazing alloy (STEMET 1108) is recommended.…”

Section: Joints High Heat Flux Testingmentioning

confidence: 99%

Development of plasma facing components for ITER at Efremov Institute

Belyakov

Filatov²,

Mazul³

17th IEEE/NPSS Symposium Fusion Engineering (Cat. No.97CH36131)

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Review of the design and R&D activity related to ITER divertor components development performed in Russia during recent years is presented. The scope of the design work and the codes used in lhis analysis are briefly reviewed. Among a variety of R&D works the key issues are presented: selection and study of armour materials, development of joining technology and mock-ups manufircturing, high heat flux tests of materials and components. The experimental and technological facilities used for this work are reviewed and a number of the alternative design solutions and experimental approaches are briefly discussed as well.

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Comparative thermal cyclic testing and strength investigation of different Be/Cu joints

Cited by 15 publications

References 5 publications

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

Establishing theoretical foundations for predicting the structural and morphological characteristics of diffusion-welded joints of the beryllium–copper composite

Development of plasma facing components for ITER at Efremov Institute

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