Effect of Large Temperature Gradients on Convective Heat Transfer: The Downstream Region

McEligot, Donald M.; Magee, P.M.; Leppert, G.

doi:10.1115/1.3689054

Cited by 69 publications

(33 citation statements)

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“…The best agreement was obtained with the correlations of Petukhov and Popov (1963) and McEligot et al (1965) for friction coefficient and Nusselt number, respectively. However, the simulations of cases 2-6 don't agree as well with the correlations as case 1.…”

Section: Friction Coefficient and Nusselt Numbersupporting

confidence: 70%

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Large eddy simulation of compressible turbulent pipe flow with heat transfer

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Section: Friction Coefficient and Nusselt Numbersupporting

confidence: 70%

“…The region of interest in this research is the "quasi-developed" region (McEligot et al, 1965), where thermal entry effects are no longer important.…”

Section: (Mceligot Et Al (1998))mentioning

confidence: 99%

Large eddy simulation of compressible turbulent pipe flow with heat transfer

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“…CFD validation calculations will be performed based on available mixed convection data [McEligot, Magee, and Leppert 1965;Perkins and McEligot 1975;Reynolds 1968;Shumway 1969;Vilemas and Poskas 1999;Bae et al 2004] to demonstrate the capability of the CFD tools to adequately calculate the appropriate fluid and heat transfer behavior. Following validation, a CFD code may be used during FY-07 to evaluate the core heat transfer in conjunction with the flow distributions in the lower and upper plenums to calculate the potential for localized hot spots in the vessel upper head and control rod apparatus (see Section 4.3.10).…”

Section: Thermal-hydraulic Design Methods Development Validation Anmentioning

confidence: 99%

“…Also, the correlations developed on the basis of available mixed convection data [McEligot, Magee, and Leppert 1965;Perkins and McEligot 1975;Reynolds 1968;Shumway 1969;Vilemas and Poskas 1999;Bae et al 2004] will be evaluated for applicability and installed in the systems analysis software.…”

Section: Thermal-hydraulic Design Methods Development Validation Anmentioning

confidence: 99%

Next Generation Nuclear Plant Research and Development Program Plan

MacDonald¹

2005

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The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years.The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents.The objectives of the NGNP Project are to:Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity.The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologiesEnergy conversion.The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen production [DOE 2004] and energy conversion technologies programs are described elsewhere.Sketch of how the NGNP might be configured to produce both electricity and hydrogen. iv Fuel Development and QualificationDevelopment and qualification of TRISO-coated low-enriched uranium fuel is a key R&D activity associated with the NGNP Program. The work is being conducted in accordance with the Technical Program Plan for the Advanced Gas Reactor Fuel Development and Qualification Program [Bell et al. 2003]. The AGR Program includes work on improving the kernel fabrication, coating, and compacting technologies, irradiation and accident testing of fuel specimens, and fuel performance and fission product transport modeling. The primary goal of these activities is to successfully demonstrate that TRISOcoated fuel can be fabricated to withstand the high temperatures, burnup, and power density requirements of a prismatic block type NGNP with an acceptable failure fraction. It is assumed that TRISO fuel that is successful in a block reactor will also be successful in pe...

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“…The empirical equation for coolant channels in pin-in block type fuels is used for the heat transfer coefficient and frication factor at the hot and cold channels (Kunitomi et al, 1991) (McEligot et al, 1965. The expression for natural convection between vertical parallel plates is used for the heat transfer coefficient at RPV outer surface (JSME, 1993).…”

Section: O Cmentioning

confidence: 99%

Design approach for mitigation of air ingress in high temperature gas-cooled reactor

Satō

Ohashi

Nakagawa

2017

Mechanical Engineering Journal

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One important safety design consideration for high temperature gas-cooled reactor (HTGR) is air ingress following a rupture of the reactor pressure boundary such as primary piping. The air intrusion to the reactor core held at high temperature through the break will results in significant oxidation of graphite components and fuels. Such oxidation may leads to the weakening of core support structures as well as fuel element damage and subsequent fission product release. This paper intends to propose a practical solution to protect the reactor from severe oxidation against air ingress accidents without reliance on subsystems. Firstly, a change is made to the center reflector structure to minimize temperature difference during the accident condition in order to reduce buoyancy-driven natural circulation in the reactor. Secondly, a modified structure of the upper reflector is suggested to prevent massive air ingress against a rupture in standpipes. As a preliminary study, a numerical analysis is performed for a typical prismatic-type HTGR to study the effectiveness of the proposed design concept using simplified lumped element models. The analysis considers internal decay heat generation and transient heat conduction from inner to outer regions at the reactor core, cooling of vessel outer surface by radiation and natural convection, and natural circulation flow in reactor. The results showed that amount of air ingress into the reactor can be significantly reduced with practical changes to local structure in the reactor.

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Effect of Large Temperature Gradients on Convective Heat Transfer: The Downstream Region

Cited by 69 publications

References 0 publications

Large eddy simulation of compressible turbulent pipe flow with heat transfer

Large eddy simulation of compressible turbulent pipe flow with heat transfer

Next Generation Nuclear Plant Research and Development Program Plan

Design approach for mitigation of air ingress in high temperature gas-cooled reactor

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