Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

McEligot, Donald M.; Condie, K. G.; McCreery, Glenn E.; McIlroy, Hugh M.; Pink, Robert J.; Hochreiter, L.E.; Jackson, J. D.; Pletcher, Richard H.; Smith, Barbara; Vukoslavčević, Petar; Wallace, J. M.; Yoo, J.Y.; Lee, J.S.; Ro, Sung Tack; Park, S.O.

doi:10.2172/911892

Cited by 7 publications

(6 citation statements)

References 47 publications

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“…The program has provided sufficiently accurate calculation for the design of previous MIR flow loop experiments, including one previous rod-bundle experiment [5]. For example, a previous rod bundle experiment which consisted of a two-rod bundle with grid spacers and enclosed in a rectangular cross-section duct which was representative of a supercritical water reactor core design.…”

Section: Rod Bundle Apparatus Designmentioning

confidence: 99%

Design of Wire-Wrapped Rod Bundle Matched Index-of-Refraction Experiments

McCreery

McIlroy

Hamman

et al. 2008

Volume 3: Thermal Hydraulics; Instrumentation and Controls

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Experiments will be conducted in the Idaho National Laboratory (INL) Matched Index-of-Refraction (MIR) Flow Facility [1] to characterize the three-dimensional velocity and turbulence fields in a wire-wrapped rod bundle typically employed in liquid-metal cooled fast reactors and to provide benchmark data for computer code validation. Sodium cooled fast reactors are under consideration for use in the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program. The experiment model will be constructed of quartz components and the working fluid will be mineral oil. Accurate temperature control (to within ± 0.05 o C) matches the index-of-refraction of mineral oil with that of quartz and renders the model transparent to the wavelength of laser light employed for optical measurements. The model will be a scaled 7-pin rod bundle enclosed in a hexagonal canister. Flow field measurements will be obtained with a LaVision 3-D particle image velocimeter (PIV) and complimented by nearwall velocity measurements obtained from a 2-D laser Doppler velocimeter (LDV). These measurements will be used as benchmark data for computational fluid dynamics (CFD) validation. The rod bundle model dimensions will be scaled up from the typical dimensions of a fast reactor fuel assembly to provide the maximum Reynolds number achievable in the MIR flow loop. A range of flows from laminar to fully-turbulent will be available with a maximum Reynolds number, based on bundle hydraulic diameter, of approximately 22,000. The fuel pins will be simulated by 85 mm diameter quartz tubes (closed on the inlet ends) and the wire-wrap will be simulated by 25 mm diameter quartz rods. The canister walls will be constructed from quartz plates. The model will be approximately 2.13 m in length. Bundle pressure losses will also be measured and the data recorded for code comparisons. The experiment design and preliminary CFD calculations, which will be used to provide qualitative hydrodynamic information, are presented in this paper.

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Section: Rod Bundle Apparatus Designmentioning

confidence: 99%

Design of Wire-Wrapped Rod Bundle Matched Index-of-Refraction Experiments

McCreery

McIlroy

Hamman

et al. 2008

Volume 3: Thermal Hydraulics; Instrumentation and Controls

View full text Add to dashboard Cite

show abstract

“…The program provided sufficiently accurate calculation for the design of the one previous rod-bundle experiment that was conducted in the MIR flow loop (McEligot, et al, 2003) and for other model configurations such as the lower plenum model discussed above. The previous rod bundle experiment was a two-rod bundle with spacers and enclosed in a rectangular cross-section duct which was representative of a supercritical water reactor core design.…”

Section: A13 Preliminary Matched-index-of-refraction Rod Bundle Expmentioning

confidence: 99%

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

McCreery¹,

McIlroy²

2007

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“…This study builds on the accomplishments of recent NERI and INERI projects led by INEEL [McEligot et al, 2002, 2003. Its unique Matched-Index-of-Refraction flow system will be utilized for benchmark measurements to assess current and future modeling techniques to benefit NGNP and VHTR programs.…”

Section: Present Randd Project For Improved Modeling and Benchmark Studmentioning

confidence: 99%

“…Under an LDRD program, ATHENA/RELAP5-3D has been linked to the Fluent CFD code to provide a tool capable of providing a macroscale flow resolution where adequate (core, piping, balance-of-plant) with a microscale flow resolution where it is necessary (inlet or outlet plenum). INEEL has recently led pertinent research projects for the NERI program ("Fundamental thermal fluid physics of high temperature flows in advanced reactors systems" ) and the Korean I-NERI program ("Advanced computational thermal fluid physics and its assessment for supercritical reactors" [McEligot et al, 2003]). These computational and experimental projects both addressed effects of fluid property variation and complex geometries, key features of flows in NGNPs and VHTRs.…”

Section: Present Randd Project For Improved Modeling and Benchmark Studmentioning

confidence: 99%

Initial Scaling Studies and Conceptual Thermal Fluids Experiments for the Prismatic NGNP Point Design

McEligot¹,

McCreery²

2004

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Table of contents iii Nomenclature iv Heated flow experiments Vertical tube Lower plenum Fluid dynamics experiments for lower plenum Concluding remarks References cited Bo* Jackson buoyancy parameter, Gr* / (Re Dh 3.425 Pr 0.8) d + wire diameter, d u / f, f friction factor, 2 b g c w / G 2 Gr* Grashof number based on heat flux, g q" wall D h 4 / (k 2) K v acceleration parameter, (b / V b 2) (dV b / dx) v Nu Nusselt number, h D h / k b Pr Prandtl number, c p /k q + heat flux, q" wall / (G c p) Re Reynolds number, 4 m / D ; Re Dh , based on hydraulic diameter, G D h / Ri overall Richardson number, g (1-2) H / (1 V b 2) Greek symbols volumetric coefficient of expansion,-(1/) (/ T) p dissipation of turbulence kinetic energy absolute viscosity kinematic viscosity, / density shear stress; w , wall shear stress Subscripts b evaluated at bulk or mixed-mean temperature (or enthalpy) D based on diameter DB Dittus-Boelter [1930] correlation Dh evaluated with hydraulic diameter f, fc forced convection i, in evaluated at inlet, entry j jet max maximum value p support post ref evaluated at reference conditions w, wall wall, evaluated at wall temperature Control Rod Drive Assembly Refueling Stand Pipe Control Rod Guide tubes

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Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

Cited by 7 publications

References 47 publications

Design of Wire-Wrapped Rod Bundle Matched Index-of-Refraction Experiments

Design of Wire-Wrapped Rod Bundle Matched Index-of-Refraction Experiments

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

Initial Scaling Studies and Conceptual Thermal Fluids Experiments for the Prismatic NGNP Point Design

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