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

McEligot, Donald M.; McCreery, Glenn E.

doi:10.2172/910959

Cited by 3 publications

(2 citation statements)

References 102 publications

(111 reference statements)

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The PIRT was a preliminary characterization of important system behavior during PCC and DCC scenarios. A study by McEligot and McCreery, 2004, identified important thermal hydraulic phenomena during normal operation and the PCC scenario for a prismatic core design. Important thermal hydraulic phenomena (including DCC phenomena) are shown schematically in Figure 4.…”

Section: Iv1 Vhtr Thermal Hydraulics Researchmentioning

confidence: 99%

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

McCreery¹,

McIlroy²

2007

View full text Add to dashboard Cite

Section: Iv1 Vhtr Thermal Hydraulics Researchmentioning

confidence: 99%

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

McCreery¹,

McIlroy²

2007

View full text Add to dashboard Cite

“…The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear 1996] is assumed in this study. VHTR vessel thermalhydraulic phenomena that are of importance during normal, reduced power, and accident operation were identified by McEligot and McCreery, 2004, and are indicated in Figure 1. The objectives of the experiments are, 1), provide benchmark data for assessment and improvement of codes proposed for NGNP designs and safety studies, and, 2), obtain a better understanding of related phenomena, behavior and needs.…”

Section: Introductionmentioning

confidence: 99%

ICONE15-10808 Scaled Experimental Modeling of VHTR Plenum Flows

McCreery

Condie

Schultz

2007

ICONE

View full text Add to dashboard Cite

The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. Various scaled heated and unheated gas and water flow facilities were investigated for modeling VHTR upper and lower plenum flows during the decay heat portion of a pressurized conduction-cooldown scenario and for modeling thermal mixing and stratification ("thermal striping") in the lower plenum during normal operation. It was concluded, based on phenomena scaling and instrumentation and other practical considerations, that a heated water flow scale model facility is preferable to a heated gas flow facility and to unheated facilities which use fluids with ranges of density to simulate the buoyancy effect of heating. For a heated water flow lower plenum model, both the Richardson numbers and Reynolds numbers may be approximately matched for conduction-cooldown natural circulation conditions. Thermal mixing during normal operation may be simulated but at lower, but still fully turbulent, Reynolds numbers than in the prototype. Natural circulation flows in the upper plenum may also be simulated in a separate heated water flow facility that uses the same plumbing as the lower plenum model. However, scaling distortions will occur due primarily to the necessity of using a reduced number of channels connected to the upper plenum than in the prototype in an otherwise geometrically scaled model.Experiments conducted in either or both facilities will meet the objectives of providing benchmark data for the validation of codes proposed for NGNP designs and safety studies, as well as providing a better understanding of the complex flow phenomena in the plenums. Power (NGNP) Project in the U.S., according to MacDonald et al., 2003. The NGNP project has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of a molten-salt coolant is also being evaluated. One of the great challenges of studying, designing, and licensing the NGNP is to verify that the intended NGNP computer code analysis tools can be used with confidence to make decisions and to assure all that the reactor systems are safe and meet their performance objectives. Comparison of computer code calculations with experimental data is the key component of code verification. This paper addresses experimental modeling of lower and upper plenum flow and thermal mixing phenomena of importance during normal operation and during a loss of forced reactor cooling (pressurized conduction cooldown) scenario. A prismatic core gas-cooled reactor following the Gas Turbine-Modular Helium Reactor (GT-MHR) design from General Atomics [GA

show abstract

Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

McCreery¹,

Condie²

2006

View full text Add to dashboard Cite

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

Cited by 3 publications

References 102 publications

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

ICONE15-10808 Scaled Experimental Modeling of VHTR Plenum Flows

Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

Contact Info

Product

Resources

About