What does this research mean for the field?

The extended CTF-SFR subchannel code provides reliable thermal-hydraulic predictions for wire-wrapped Sodium Fast Reactor assemblies, accurately reproducing pressure losses and temperature profiles, despite limitations in modeling axial conduction. Novelty: ClaimNovelty.METHODOLOGICAL. Consensus alignment: ConsensusAlignment.NEUTRAL.

What question did this study set out to answer?

To evaluate and validate the extended CTF subchannel code for thermal-hydraulic modeling of Sodium Fast Reactors (SFRs) incorporating liquid metal properties and wire-wrapped hexagonal geometries.

August 23, 2026Nuclear Technology

Validation of Subchannel Code CTF for Sodium Fast Reactor Modeling

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Authors

JVJonas Vanderwaeren HVHéloïse Velardo CSChristophe Schneidesch

Discussion

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Overview

Benchmark validation study demonstrates reliable thermal-hydraulic predictions by the CTF code in sodium fast reactors, highlighting its capability for wire-wrapped fuel assemblies.

Key Points

To evaluate and validate the extended CTF subchannel code for thermal-hydraulic modeling of Sodium Fast Reactors (SFRs) incorporating liquid metal properties and wire-wrapped hexagonal geometries.
Integrated low-Prandtl-number fluid models, sodium heat transfer correlations, and hexagonal wire-wrapped geometry capabilities into the CTF subchannel code.
Validated the tool against isothermal pressure drop data from the Texas A&M University 61-pin wire-wrapped bundle and multi-condition heat balance data from the SEFOR facility.
Performed code-to-code comparisons against TrioMC and conducted sensitivity studies on mixing and conduction models.
CTF-SFR accurately reproduced experimental frictional pressure losses across the full range of Reynolds numbers in the TAMU 61-pin bundle.
The simulation matched 61 heat balance points in SEFOR single-assembly tests, showing close agreement with reference average and peak sodium temperatures.
Comparisons with TrioMC highlighted discrepancies in axial temperature profiles due to unmodeled axial conduction, as well as overpredicted corner channel peak temperatures in bare rod bundles.

Cite This Study

Vanderwaeren et al. (2026) studied this question.

synapsesocial.com/papers/6a8aad667677a3411444586d https://doi.org/10.1080/00295450.2026.2712091

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August 23, 2026Nuclear Technology

Validation of Subchannel Code CTF for Sodium Fast Reactor Modeling

View Full Paper

Ask AI

Bookmark

Authors

JVJonas Vanderwaeren HVHéloïse Velardo CSChristophe Schneidesch

Discussion

Member takes

Overview

Benchmark validation study demonstrates reliable thermal-hydraulic predictions by the CTF code in sodium fast reactors, highlighting its capability for wire-wrapped fuel assemblies.

Key Points

To evaluate and validate the extended CTF subchannel code for thermal-hydraulic modeling of Sodium Fast Reactors (SFRs) incorporating liquid metal properties and wire-wrapped hexagonal geometries.
Integrated low-Prandtl-number fluid models, sodium heat transfer correlations, and hexagonal wire-wrapped geometry capabilities into the CTF subchannel code.
Validated the tool against isothermal pressure drop data from the Texas A&M University 61-pin wire-wrapped bundle and multi-condition heat balance data from the SEFOR facility.
Performed code-to-code comparisons against TrioMC and conducted sensitivity studies on mixing and conduction models.
CTF-SFR accurately reproduced experimental frictional pressure losses across the full range of Reynolds numbers in the TAMU 61-pin bundle.
The simulation matched 61 heat balance points in SEFOR single-assembly tests, showing close agreement with reference average and peak sodium temperatures.
Comparisons with TrioMC highlighted discrepancies in axial temperature profiles due to unmodeled axial conduction, as well as overpredicted corner channel peak temperatures in bare rod bundles.

Cite This Study

Vanderwaeren et al. (2026) studied this question.

synapsesocial.com/papers/6a8aad667677a3411444586d https://doi.org/10.1080/00295450.2026.2712091

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