Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

Urgorri, Fernando Roca; Moreno, Carlos; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, Iván; Palermo, Iole; Ibarra, Á.

doi:10.1088/1741-4326/aa7f9d

Cited by 18 publications

(17 citation statements)

References 24 publications

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“…There is one order of magnitude of difference in tritium concentration in side and Hartmann gaps. The steady-state permeation rate, from the gap flow into the helium channels, is around six times higher than the value predicted by a system-level model that considers an evenly distributed flow rate through the gaps [129]. This comparison highlights the need for 3D transport models to predict tritium distribution in liquid metal blankets in order to reduce uncertainties and inaccuracy caused by assumptions, geometrical approximations and simplifications of the physics, as used in system models.…”

Section: Wcll Demo Blanketmentioning

confidence: 86%

“…The obtained velocity profile has been used as input for a 3D tritium transport model of the complete annular channel between external wall and FCI. This model considers an exponentially decreasing volumetric tritium generation along the radial direction and constant transport properties [129]. Figure 15 (right) displays concentration contours in the midsection of the channel.…”

Section: Wcll Demo Blanketmentioning

confidence: 99%

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MHD R&D Activities for Liquid Metal Blankets

et al. 2021

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According to the most recently revised European design strategy for DEMO breeding blankets, mature concepts have been identified that require a reduced technological extrapolation towards DEMO and will be tested in ITER. In order to optimize and finalize the design of test blanket modules, a number of issues have to be better understood that are related to the magnetohydrodynamic (MHD) interactions of the liquid breeder with the strong magnetic field that confines the fusion plasma. The aim of the present paper is to describe the state of the art of the study of MHD effects coupled with other physical phenomena, such as tritium transport, corrosion and heat transfer. Both numerical and experimental approaches are discussed, as well as future requirements to achieve a reliable prediction of these processes in liquid metal blankets.

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Section: Wcll Demo Blanketmentioning

confidence: 86%

Section: Wcll Demo Blanketmentioning

confidence: 99%

MHD R&D Activities for Liquid Metal Blankets

et al. 2021

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“…FUS-TPC, developed by Franza et al at ENEA, Italy in 2011 is a fusion version of the tritium permeation code for sodiumcooled fast reactors (SFR-TPC) [133]. CIEMAT has led the development of a tritium transport toolkit for EcosimPro, and applied to several of the EU candidate DEMO blanket concepts [134]. The fusion reactor tritium analysis system (TAS) was developed in the 2000s by the team for Frontier Development of Science (FDS) team at the Institute of Nuclear Energy Safety Technology, China [135].…”

Section: Tritium Permeation and Managementmentioning

confidence: 99%

Physics and technology considerations for the deuterium–tritium fuel cycle and conditions for tritium fuel self sufficiency

et al. 2020

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The tritium aspects of the DT fuel cycle embody some of the most challenging feasibility and attractiveness issues in the development of fusion systems. The review and analyses in this paper provide important information to understand and quantify these challenges and to define the phase space of plasma physics and fusion technology parameters and features that must guide a serious R&D in the world fusion program. We focus in particular on components, issues and R&D necessary to satisfy three ‘principal requirements’: (1) achieving tritium self-sufficiency within the fusion system, (2) providing a tritium inventory for the initial start-up of a fusion facility, and (3) managing the safety and biological hazards of tritium. A primary conclusion is that the physics and technology state-of-the-art will not enable DEMO and future power plants to satisfy these principal requirements. We quantify goals and define specific areas and ideas for physics and technology R&D to meet these requirements. A powerful fuel cycle dynamics model was developed to calculate time-dependent tritium inventories and flow rates in all parts and components of the fuel cycle for different ranges of parameters and physics and technology conditions. Dynamics modeling analyses show that the key parameters affecting tritium inventories, tritium start-up inventory, and tritium self-sufficiency are the tritium burn fraction in the plasma (f b), fueling efficiency (η f), processing time of plasma exhaust in the inner fuel cycle (t p), reactor availability factor (AF), reserve time (t r) which determines the reserve tritium inventory needed in the storage system in order to keep the plant operational for time t r in case of any malfunction of any part of the tritium processing system, and the doubling time (t d). Results show that η f f b > 2% and processing time of 1–4 h are required to achieve tritium self-sufficiency with reasonable confidence. For η f f b = 2% and processing time of 4 h, the tritium start-up inventory required for a 3 GW fusion reactor is ∼11 kg, while it is <5 kg if η f f b = 5% and the processing time is 1 h. To achieve these stringent requirements, a serious R&D program in physics and technology is necessary. The EU-DEMO direct internal recycling concept that carries fuel directly from the plasma exhaust gas to the fueling systems without going through the isotope separation system reduces the overall processing time and tritium inventories and has positive effects on the required tritium breeding ratio (TBRR). A significant finding is the strong dependence of tritium self-sufficiency on the reactor availability factor. Simulations show that tritium self-sufficiency is: impossible if AF < 10% for any η f f b, possible if AF > 30% and 1% ⩽ η f f b ⩽ 2%, and achievable with reasonable confidence if AF > 50% and η f f b > 2%. These results are of particular concern in light of the low availability factor predicted for the near-term plasma-based experimental facilities (e.g. FNSF, VNS, CTF), and can have repercussions on tritium economy in DEMO reactors as well, unless significant advancements in RAMI are made. There is a linear dependency between the tritium start-up inventory and the fusion power. The required tritium start-up inventory for a fusion facility of 100 MW fusion power is as small as 1 kg. Since fusion power plants will have large powers for better economics, it is important to maintain a ‘reserve’ tritium inventory in the tritium storage system to continue to fuel the plasma and avoid plant shutdown in case of malfunctions of some parts of the tritium processing lines. But our results show that a reserve time as short as 24 h leads to unacceptable reserve and start-up inventory requirements. Therefore, high reliability and fast maintainability of all components in the fuel cycle are necessary in order to avoid the need for storing reserve tritium inventory sufficient for continued fusion facility operation for more than a few hours. The physics aspects of plasma fueling, tritium burn fraction, and particle and power exhaust are highly interrelated and complex, and predictions for DEMO and power reactors are highly uncertain because of lack of experiments with burning plasma. Fueling by pellet injection on the high field side of tokamak has evolved to be the preferred method to fuel a burning plasma. Extrapolation from the DIII-D penetration scaling shows fueling efficiency expected in DEMO to be <25%, but such extrapolations are highly uncertain. The fueling efficiency of gas in a reactor relevant regime is expected to be extremely poor and not very useful for getting tritium into the core plasma efficiently. Gas fueling will nonetheless be useful for feedback control of the divertor operating parameters. Extensive modeling has been carried out to predict burn fraction, fueling requirements, and fueling efficiency for ITER, DEMO, and beyond. The fueling rate required to operate Q = 10 ITER plasmas in order to provide the required core fueling, helium exhaust and radiative divertor plasma conditions for acceptable divertor power loads was calculated. If this fueling is performed with a 50–50 DT mix, the tritium burn fraction in ITER would be ∼0.36%, which is too low to satisfy the self-sufficiency conditions derived from the dynamics modeling for fusion reactors. Extrapolation to DEMO using this approach would also yield similarly low burn fraction. Extensive analysis presented shows that specific features of edge neutral dynamics in ITER and fusion reactors, which are different from present experiments, open possibilities for optimization of tritium fueling and thus to improve the burn fraction. Using only tritium in pellet fueling of the plasma core, and only deuterium for edge density, divertor power load and ELM control results in significant increase of the burn fraction to 1.8–3.6%. These estimates are performed with physics models whose results cannot be fully validated for ITER and DEMO plasma conditions since these cannot be achieved in present tokamak experiments. Thus, several uncertainties remain regarding particle transport and scenario requirements in ITER and DEMO. The safety standard requirements for protection of the public and release guidelines for tritium have been reviewed. General safety approaches including minimizing tritium inventories, reducing tritium permeation through materials, and decontaminating material for waste disposal have been suggested.

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“…In order to extract most of the reactor power, this blanket has to manage a huge amount of PbLi which leads to this high velocity (~2 cm s −1 in the blanket [2]). An immediate consequence is that the tritium concentration in the liquid is rather low, meaning that the tritium partial pressure in DCLL is about two orders of magnitude lower than the expected one in other blanket concepts [3,4]. From the point of view of safety, it can be an advantage since the permeation of tritium to the secondary coolant is reduced.…”

Section: Introductionmentioning

confidence: 99%

The tritium extraction and removal system for the DCLL-DEMO fusion reactor

Garcinuño

Rapisarda

Antunes³

et al. 2018

Nucl. Fusion

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During the pre-conceptual design phase of DEMO, different alternatives have been explored to be implemented as tritium extraction and removal system (TERS) for the blanket concepts considered in EUROfusion. The TERS is conceived to extract tritium from the breeder and to route it to the Tritium Plant for final processing. A careful review showed that those blankets operated with PbLi should use the permeation against vacuum (PAV) technique as primary option which is based on a one-step, fully continuous procedure. In this paper, a conceptual design of the TERS for the dual coolant lithium lead (DCLL) breeding blanket is presented, based on the European DEMO2015 layout (18 sectors, 2037 MW fusion power). The P&ID of the proposed TERS, integrated in the DCLL-PbLi loop, includes valves and instrumentation, as well as a revised design of the DCLL-PAV. The dimensioning of the permeator considered a tritium extraction efficiency of 80%. An exhaustive investigation on the vacuum system needed for the PAV is also presented. The choice of the most promising vacuum systems took into account the reliability and tritium compatibility of both high and rough pumps. Their pumping requirements, which are dependent on the PAV efficiency, tritium solubility and tritium partial pressure in the loop, are also discussed in this work.

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Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

Cited by 18 publications

References 24 publications

MHD R&D Activities for Liquid Metal Blankets

MHD R&D Activities for Liquid Metal Blankets

Physics and technology considerations for the deuterium–tritium fuel cycle and conditions for tritium fuel self sufficiency

The tritium extraction and removal system for the DCLL-DEMO fusion reactor

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