The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

Grishchenko, Dmitry; Jeltsov, Marti; Kööp, Kaspar; Karbojian, Aram; Villanueva, Walter; Kudinov, Pavel

doi:10.1016/j.nucengdes.2014.11.045

Cited by 43 publications

(8 citation statements)

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“…were benchmarked to evaluate their extended function for lead/lead‐bismuth system 7‐9,39 . Later, with the purpose to validate multidimensional CFD/STH codes, a third leg with a 3D test section was added to the original TALL facility 41 . More detailed information of TALL and TALL‐3D design can be seen in References 7, 8, 39–41.…”

Section: Benchmarking Of Lethac With Tall Facilitymentioning

confidence: 99%

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

Wei

Wang

et al. 2021

Int J Energy Res

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Summary In this paper, development and preliminary validation of an in‐house code LETHAC (lead‐based fast reactor thermal‐hydraulic analysis code) is introduced. LETHAC is developed for predicting the thermal‐hydraulic transient performance of a lead‐bismuth reactor. The mathematical models and module relations of LETHAC are presented in detail. The most‐used LBE (lead‐bismuth eutectic) flow and heat transfer correlations are compared against data from KYLIN‐II 61‐rod wire‐wrapped bundle experiments. Results show little difference between the chosen heat transfer correlations as the relative error of these correlations is all within 2%. Novendstern correlation is recommended for wire‐wrapped bundle frictional resistance coefficient calculation as it fits the experimental data best. Besides, the transient calculation capability of LETHAC is also preliminarily evidenced by TALL transient tests. Good agreements are shown between the calculated temperatures and the experiment ones. Relative errors between calculation results and experimental data are almost within 10% during the whole transient process, which indicates LETHAC can be used for predicting the transient behavior of lead/lead‐bismuth cooled systems.

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Section: Benchmarking Of Lethac With Tall Facilitymentioning

confidence: 99%

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

Wei

Wang

et al. 2021

Int J Energy Res

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“…In contrast, the ability of CFD to simulate geometrical effects directly makes it easier to argue that experiments performed for other designs (and thus in different geometrical configurations) will provide a useful contribution for validating the new tool. Thanks to these arguments, multi-scale/CFD tool may thus rely on a growing experimental database for validating coupled effects: for the effect of pool thermal-hydraulics on natural convection, TALL-3D [25] (SET), CIRCE-HERO [26] (IET) and E-SCAPE [27] (IET) may be considered; for the effect of inter-wrapper flow on S/A cooling, THEADES [28] (SET), PLANDTL-1 [29] (IET) and PLANDTL-2 [30] (IET) are available or will be in the near future; -CLEAR-S [31] (IET) will present an interesting case where it will be possible to validate the effect of pool TH and inter-wrapper flow on global natural convection at the same time; for integral validation, PHENIX [24], EBR-II [32] and FFTF [33] tests are or will be available in the short-term.…”

Section: Validation Stepmentioning

confidence: 99%

VVUQ of a thermal-hydraulic multi-scale tool on unprotected loss of flow accident in SFR reactor

Marie

Marrel

et al. 2021

EPJ Nuclear Sci. Technol.

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Within the framework of the French 4th-generation Sodium-cooled Fast Reactor safety assessment, methodology on VVUQ (Verification, Validation, Uncertainty Quantification) is conducted to demonstrate that the CEA's thermal-hydraulic Scientific Computation Tools (SCTs) are effective and operational for design and safety studies purposes on this type of reactor. This VVUQ-based qualification is a regulatory requirement from the French Nuclear Safety Authority (NSA). In this paper, the current practice of VVUQ approach application for a SFR accidental transient is described with regard to the NSA requirements. It constitutes the first practical, progressively improvable approach. As the SCT is qualified for a given version on a given scenario, the transient related to a total unprotected station blackout has been selected. As it is a very complex multi-scale transient, the SCT MATHYS (which is a coupling of the CATHARE2 tool at system scale, TrioMC tool at component scale and TrioCFD tool at local scale) is used. This paper presents the preliminary VVUQ application to the qualification of this tool on this selected transient. In addition, this work underlines some feedback on design and R&D aspects that should be addressed in the future to improve the SCT.

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“…5). Thermal stratification and mixing phenomena were investigated in small scale apparatus like the TALL-3D facility [29] (Thermal-hydraulic Lead-bismuth Loop with 3D flow test section) developed at KTH (Royal Institute of Technology, Stockholm, Sweden). Solidification/remelting in buoyancy driven lead flow was performed in the SESAME-stand experimental facility by CVR (Research Centre Rez, Czech Republic).…”

Section: Pool Thermal Hydraulicsmentioning

confidence: 99%

SESAME project: advancements in liquid metal thermal hydraulics experiments and simulations

Tarantino

Roelofs

Shams

et al. 2020

EPJ Nuclear Sci. Technol.

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Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Sodium and Liquid lead (-alloys) are considered the short and long term solution respectively, as coolant in GEN-IV reactor. Thermal-hydraulics of liquid metals plays a key role in the design and safety assessments of these reactors. Therefore, this is the main topic of a large European collaborative program (the Horizon 2020 SESAME) sponsored by the European Commission. This paper will present the progress in the project with respect to liquid metal cooled reactor thermal-hydraulics (liquid metal heat transfer, fuel assembly thermal-hydraulics, pool thermal-hydraulics, and system thermal-hydraulics). New reference data, both experimental and high-fidelity numerical data is being generated. And finally, when considering the system scale, the purpose is to validate and improve system thermal-hydraulics models and codes, but also to further develop and validate multi-scale approaches under development.

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The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

Cited by 43 publications

References 0 publications

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

VVUQ of a thermal-hydraulic multi-scale tool on unprotected loss of flow accident in SFR reactor

SESAME project: advancements in liquid metal thermal hydraulics experiments and simulations

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