Pressure Measurements in a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle

Vaghetto, Rodolfo; Jones, P.; Goth, Nolan; Childs, Mason; Lee, Saya; Nguyen, Diep Thi Ngoc; Hassan, Yassin A.

doi:10.1115/1.4038086

Cited by 31 publications

(6 citation statements)

References 20 publications

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“…Often however, the uncertainties of pressure drop measurements associated with these experiments is high due to the geometrical complexity of the hexagonal wire-wrapped fuel bundle. Recently [23,24], a database of pressure drops and flow-field measurements in a 61-pin wire-wrapped hexagonal fuel bundle was developed with the sole purpose to benchmark the existing correlations and validate the CFD calculations. These experiments investigate the flow characteristics in the near-wall region of the 61-pin wire-wrapped hexagon fuel bundle.…”

Section: Challenges and Opportunities For Enhanced Computation Capacity And Extended Investigation In The Sodium-cooled Fast Reactors Fiementioning

confidence: 99%

International Benchmark Activity in the Field of Sodium Fast Reactors

Luca¹,

Pasquale²,

Cherubini³

et al. 2021

Recent Advances in Numerical Simulations

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Global interest in fast reactors has been growing since their inception in 1960 because they can provide efficient, safe, and sustainable energy. Their closed fuel cycle can support long-term nuclear power development as part of the world’s future energy mix and decrease the burden of nuclear waste. In addition to current fast reactors construction projects, several countries are engaged in intense R&D and innovation programs for the development of innovative, or Generation IV, fast reactor concepts. Within this framework, NINE is very actively participating in various Coordinated Research Projects (CRPs) organized by the IAEA, aimed at improving Member States’ fast reactor analytical simulation capabilities and international qualification through code-to-code comparison, as well as experimental validation on mock-up experiment results of codes currently employed in the field of fast reactors. The first CRP was focused on the benchmark analysis of Experimental Breeder Reactor II (EBR-II) Shutdown Heat Removal Test (SHRT-17), protected loss-of-flow transient, which ended in the 2017 with the publication of the IAEA-TECDOC-1819. In the framework of this project, the NINE Validation Process– developed in the framework of NEMM (NINE Evaluation Model Methodology) – has been proposed and adopted by most of the organizations to support the interpretation of the results calculated by the CRP participants and the understanding of the reasons for differences between the participants’ simulation results and the experimental data. A second project regards the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the Fast Flux Test Facility (FFTF), the Loss of Flow Without Scram (LOFWOS) Test #13, started in 2018. A detailed nodalization has been developed by NINE following its nodalization techniques and the NINE validation procedure has been adopted to validate the Simulation Model (SM) against the experimental data of the selected test. The third activity deals with the neutronics benchmark of China Experimental Fast Reactor (CEFR) Start-Up Tests, a CRP proposed by the China Institute of Atomic Energy (CIAE) launched in 2018 the main objective of which is to improve the understanding of the start-up of a SFR and to validate the fast reactor analysis computer codes against CEFR experimental data. A series of start-up tests have been analyzed in this benchmark and NINE also proposed and organized a further work package focused on the sensitivity and uncertainty analysis of the first criticality test. The present chapter intends to summarize the results achieved using the codes currently employed in the field of fast reactor in the framework of international projects and benchmarks in which NINE was involved and emphasize how the application of developed procedures allows to validate the SM results and validate the computer codes against experimental data.

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Section: Challenges and Opportunities For Enhanced Computation Capacity And Extended Investigation In The Sodium-cooled Fast Reactors Fiementioning

confidence: 99%

International Benchmark Activity in the Field of Sodium Fast Reactors

Luca¹,

Pasquale²,

Cherubini³

et al. 2021

Recent Advances in Numerical Simulations

View full text Add to dashboard Cite

show abstract

“…Next to that, they claim that the shape and magnitude of the mean horizontal velocity components are in satisfactory agreement. While friction data generated in the 61-pin bundle (Vaghetto et al, 2018) have been compared with RANS simulations but not yet published, this is still an integral component of the data production program.…”

Section: Data Production Programmentioning

confidence: 99%

Towards validated prediction with RANS CFD of flow and heat transport in a wire-wrap fuel assembly

Roelofs

Uitslag-Doolaard

Dovizio

et al. 2019

Nuclear Engineering and Design

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Liquid metal fast reactors (LMFRs) are foreseen to play an important role in the future of nuclear energy, thanks to their increased fuel utilization and safety features profiting from the optimal heat transfer performance of the metallic coolants. Accurate thermal-hydraulic analysis of their fuel assemblies, typically employed with wire-wraps as spacers, is recognized as a crucial scientific and engineering contribution to support the deployment of such technology. This challenges the modeling and simulation community. To this aspect, various reference databases (both experimental and numerical) for different wire-wrapped fuel assembly configurations have been created recently and are being used for validation of engineering simulation approaches based on Reynolds Averaged Navier Stokes (RANS) modelling. These databases include: • 7-pin rod bundle: A detailed experiment with Particle Image Velocimetry (PIV) is performed. In order to allow accurate measurements of the flow topology, a matched-index-of-refraction technique was used employing water as working fluid. • 19-pin bundle: A series of experiments is performed covering a wide range of Reynolds and Peclet numbers as well as thermal powers. The experiments use liquid lead-bismuth eutectic as working fluid. The measurements include pressure drop and local temperatures. • 61-pin rod bundle: This large eddy simulation including conjugate heat transfer from the pin cladding to the coolant allows to bridge the gap from small bundles (less than 37 pins) to large bundles (more than 37 pins). In literature, a fundamental different behavior has been observed for small bundles compared to large bundles. • 127-pin bundle: Isothermal experiments using lead-bismuth eutectic characterizing pressure drop are performed on a full scale fuel assembly representative for the MYRRHA reactor. • Infinite pin bundle: This reference quasi-direct numerical simulation profits from periodicity in all directions. It provides a detailed view into the flow field and in addition reveals details of the heat transfer from the rod bundle into the flow. Reference databases aim to serve the nuclear scientific community to validate engineering simulation approaches. The paper will introduce these reference databases, and how they have been used to validate RANS based turbulence modelling approaches within a mainly European context.

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“…In previous decades, researchers have experimentally and numerically investigated the hydraulic characteristics in a wire-wrapped fuel bundle channel on the basis of dimensional analysis and similarity theory. Experimental research: Vaghetto et al [ 10 ] used p-Cymene as the working fluid (medium) and investigated the pressure drop in a 61-pin wire-wrapped fuel assembly with a lead pitch of 476 mm. They found that the detailed Cheng and Todreas correlation [ 11 ] provided the most accurate prediction of pressure drop.…”

Section: Introductionmentioning

confidence: 99%