Engineering design of a Permeator Against Vacuum mock-up with niobium membrane

Papa, Francesca; Utili, Marco; Venturini, Alessandro; Caruso, Gianfranco; Savoldi, Laura; Bonifetto, Roberto; Valerio, Domenico; Allio, Andrea; Collaku, Aldo; Tarantino, M.

doi:10.1016/j.fusengdes.2021.112313

Cited by 19 publications

(11 citation statements)

References 21 publications

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“…Permeator against vacuum [48][49][50], liquid-liquid contractor [51], gas-bubble columns [52], vacuum sieve tray [53].…”

Section: Tritium Extraction System (Tes)mentioning

confidence: 99%

“…Tritium extraction from a liquid lithium carrier is better understood. Tritium extraction from PbLi has been investigated in depth for EU-DEMO [48][49][50], providing possible values for η TES . Therefore, a detailed two-parameter sensitivity analysis (τ BZ and η TES ) is performed in this paper for STEP-Li to account for uncertainties in the estimation of TBR r and I startup .…”

Section: Step-limentioning

confidence: 99%

“…Our analysis accounts for this by exploring a much wider range of possible τ BZ, from 1 h to 240 h. As for the TES, two possible technologies are considered: permeator against vacuum (PAV) and liquid-liquid extraction (LLE). The PAV has been extensively investigated for the DEMO designs using PbLi breeder [48][49][50]. The estimated η TES spans from 3% to 80% in those works, resulting in a second large source of uncertainty.…”

Section: Step-li Fcmentioning

confidence: 99%

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Modeling and analysis of the tritium fuel cycle for ARC- and STEP-class D-T fusion power plants

Meschini,

Ferry,

Delaporte-Mathurin

et al. 2023

Nucl. Fusion

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The limited tritium resources available for the first fusion power plants (FPPs) make fuel self-sufficiency and tritium inventory minimization leading issues in FPP design. This work builds on the model proposed by M. Abdou et al. \cite{abdou2020physics}, which analyzed the fuel cycle of a DEMO-class FPP with a time-dependent system-level model. Here, we use a modified version of their model to analyze the fuel cycle of an ARC-class tokamak and two versions of a STEP-class tokamak. The ARC-class tokamak breeds tritium in a FLiBe liquid immersion blanket (LIB), while the STEP-class tokamak breeds tritium utilizing either a liquid-lithium blanket design or an Encapsulated Breeding Blanket (EBB). A time-dependent system-level model is developed in Matlab Simulink~\textregistered\hspace{.2em} to simulate the evolution of tritium flows and tritium inventories in the fuel cycle. The main goals of this work are to assess tritium self-sufficiency of the ARC- and STEP-class designs and to determine quantitative design requirements that can be used to analyze the adequacy of a proposed fuel cycle system. These design requirements are aimed at achieving a low tritium inventory doubling time ($t_d$) and a low start-up inventory ($I_{\mathrm{startup}}$) while keeping the required tritium breeding ratio (TBR$_r$) as low as possible. We also consider how improvements in fuel cycle technology and plasma operations affect TBR$_r$ and $I_{\mathrm{startup}}$. The model results show that TBR$_r$ for ARC- and STEP-class FPPs should be achievable if the tritium burn efficiency (TBE) reaches 0.5-1\% (TBR$_r <$ 1.2). This assumes significant, but attainable, improvements over current abilities. However, the model results indicate that a FPP must achieve ambitious performance targets, including FPP availability $>$70\%, tritium processing time $<$4~h, and the implementation of direct internal recycling. If future research yields major improvements to achievable TBE, it may be possible to achieve tritium self-sufficiency while operating at lower availability and without implementing DIR.

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“…Permeator against vacuum [48][49][50], liquid-liquid contractor [51], gas-bubble columns [52], vacuum sieve tray [53].…”

Section: Tritium Extraction System (Tes)mentioning

confidence: 99%

Section: Step-limentioning

confidence: 99%

Section: Step-li Fcmentioning

confidence: 99%

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Modeling and analysis of the tritium fuel cycle for ARC- and STEP-class D-T fusion power plants

Meschini,

Ferry,

Delaporte-Mathurin

et al. 2023

Nucl. Fusion

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“…The PAV technology is based on the phenomenon of tritium permeation through a membrane towards a secondary side where vacuum is pumped. In this way, a pressure gradient is established promoting tritium extraction [8]. Finally, in the LVC the liquid metal flows from an upper chamber to a bottom vacuumed one; these chambers are separated by a tray equipped with nozzles of diameter of the order of the millimetre, which allows the alloy to form an unstable liquid jet of droplets.…”

Section: Introductionmentioning

confidence: 99%

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

Utili¹,

Alberghi²,

Candido³

et al. 2022

Nucl. Fusion

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The experimental qualiﬁcation of the Tritium Extraction Unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the Water-Cooled Lithium-Lead (WCLL) breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction eﬃciency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone research centre, Italy. The facility is able to qualify Gas-Liquid Contactor (GLC), Permeator Against Vacuum (PAV) and Liquid-Vacuum Contactor (LVC) technologies at diﬀerent temperatures, lithium-lead mass ﬂow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extracted from the liquid metal in the GLC mock-up, which uses pure helium or a mixture constituted by helium and hydrogen as stripping gas and works in the temperature range between 300 and 500 °C.

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“…The candidate materials for membranes are niobium and vanadium [3]. Although mock-ups of the Tritium Extraction Unit are under manufacturing with both niobium [4] [5] and vanadium [6] [4] as membrane materials, their compatibility with LiPb was studied in stagnant conditions by Feuerstein et al [7] and Grabner et al [8]: they deemed both elements to be very stable in LiPb up to 650°C. The solubility of V is reported to be 1.1 and 3.5 ppmw at 500°C and 600°C respectively, whereas the solubility of Nb is reported to be 0.053 ppmw at 600°C [7].…”

Section: Introductionmentioning

confidence: 99%

Compatibility of niobium, vanadium and P22 steel in high temperature flowing LiPb

Venturini

Bassini

Ciantelli

et al. 2022

Journal of Nuclear Materials

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Engineering design of a Permeator Against Vacuum mock-up with niobium membrane

Cited by 19 publications

References 21 publications

Modeling and analysis of the tritium fuel cycle for ARC- and STEP-class D-T fusion power plants

Modeling and analysis of the tritium fuel cycle for ARC- and STEP-class D-T fusion power plants

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

Compatibility of niobium, vanadium and P22 steel in high temperature flowing LiPb

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