Development of an object-oriented, thermal-hydraulics model for ARC FLiBe loop safety assessment

Meschini, Samuele; Testoni, Raffaella; Zucchetti, Massimo

doi:10.1016/j.fusengdes.2022.113095

Cited by 3 publications

(2 citation statements)

References 23 publications

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“…It also provides efficient heat conduction away from the coupled first wall (FW)-VV structure, which is very important at the high power density ARC proposes. Analyses of ARC have been carried out in the fields of materials science [10][11][12], thermal analysis [13][14][15], neutronics [13,[16][17][18], and plasma physics [2,13,19].…”

Section: Introductionmentioning

confidence: 99%

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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Section: Introductionmentioning

confidence: 99%

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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“…Conversely, other system-level tools for the dynamic simulation of liquid metal or molten salt BB loops already exist, e.g. a code developed with the EcosimPro simulation platform including the tritium transport modelling in the PbLi loop of the HCLL-TBS [11], or a thermal-hydraulic model of the FLiBe loop of the ARC reactor developed with Modelica [12], or models to predict MHD pressure drops in PbLi implemented in the system-level thermal-hydraulic code RELAP5/MOD3.3 [13]. However, to the best of the authors' knowledge, a system-level analysis on the generation and transport of ACPs in the WCLL PbLi loop was yet to be performed.…”

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Modeling the Transport of Activated Corrosion Products in the WCLL PbLi Loop for ITER and the EU DEMO With the GETTHEM Code

et al. 2023

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This work presents the results of the implementation in the GEneral Tokamak THErmalhydraulic Model of available models for the generation and transport of any dispersed material in the PbLi eutectic mixture circulating in the Water-Cooled Lithium-Lead Breeding Blanket of the EU DEMO fusion reactor, with main focus on Activated Corrosion Products as solid suspension. A simple test case is used to show that the distribution of the concentration of activated corrosion products at any point of the PbLi loop, both in the water-cooled lithium-lead breeding blanket and in the related ITER Test Blanket System, can be determined by the model. The results obtained with this tool can be useful not only for radiological safety purposes, but also because activated corrosion products may affect the PbLi flow itself and the efficiency of the tritium removal system, with consequences on the achievable Tritium Breeding Ratio. A rigorous verification of the model is also performed. INDEX TERMS Activated corrosion products, blanket, ITER, modelling, nuclear fusion I. INTRODUCTIONThe design of the EU DEMO, performed by the EUROfusion Consortium [1], takes advantage of different computational tools, often aimed at modelling single physics and/or single systems. However, system-level tools used for plant integration studies require different pieces of physics and/or different subsystems to be modelled at the same time; this is one of the aims of the system-level GEneral Tokamak THErmal-hydraulic Model (GETTHEM), developed at Politecnico di Torino since 2015 [2], for the thermal-hydraulic modelling of the Breeding Blanket (BB) and related subsystems, e.g. the Primary Heat Transfer System (PHTS) and the Balance-of-Plant. The code was applied in the past for the thermal-hydraulic transient analysis of the Helium-Cooled Pebble Bed and Water-Cooled Lithium-Lead (WCLL) [3] BB designs in normal and off-normal scenarios [4,5]. In recent years, a model of the WCLL PbLi loop was introduced in GETTHEM, including MHD pressure drops [6], but a comprehensive model of the PbLi loop must include also additional models relevant to plant operation and safetye.g. for tritium production, transport, and extraction, as well as for the transport of other radioactive elements such as the Activated Corrosion Products (ACPs)since the different phenomena involved are interdependent with each other, hence a self-consistent model should be able to address them at the same time. The need for a self-consistent assessment of this phenomenon with a multiphysics approach is also highlighted by the recent efforts to couple the OSCAR tool for the assessment of the ACPs with other codes such as MCNP [7] or RAVEN [8], which are being pursued for the water cooling loops of the PHTS of tokamaks, but is still missing for liquid metal systems [9].

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Impact Assessment of Radiative Heat Transport in ARC-Class Reactor FLiBe Liquid Immersion Blanket

Ferrero

Testoni

Zucchetti

2023

Nuclear Science and Engineering

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Development of an object-oriented, thermal-hydraulics model for ARC FLiBe loop safety assessment

Cited by 3 publications

References 23 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

Modeling the Transport of Activated Corrosion Products in the WCLL PbLi Loop for ITER and the EU DEMO With the GETTHEM Code

Impact Assessment of Radiative Heat Transport in ARC-Class Reactor FLiBe Liquid Immersion Blanket

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