Next Generation Nuclear Plant Methods Technical Program Plan

Schultz, Richard R.; Ougouag, Abderrafi M.; Nigg, David W.; Gougar, Hans D.; Johnson, Richard W.; Terry, William S.; Oh, Chang Ho; McEligot, Donald W.; Johnsen, G.W.; McCreery, Glenn E.; Yoon, Woo Young; Sterbentz, James W.; Herring, J. Steve; Taiwo, T. A.; Wei, Thomas Y. C.; Pointer, W. D.; Yang, Won Seok; Farmer, Michael T.; Khalil, Hassan A.; Feltus, Madeline Anne

doi:10.2172/911934

Cited by 15 publications

(15 citation statements)

References 16 publications

(19 reference statements)

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“…However, during the interim, first-cut PIRTs have been used instead as a guide for the initial R&D work and planning for both block-type and pebble-bed-type gas-cooled reactors. The first-cut PIRTs are based on observations from seasoned gas-cooled reactor experts and engineering judgment; these factors were used by a team assembled to define the first PIRT for the prismatic and pebble-bed reactors and described in Appendix A of Schultz et al [2008] and is documented in detail in Lee, Wei, and Schultz et al [2005] and Ball et al [2008]. The results of the "first-cut" PIRTs for steady-state operation, PCC, and DCC scenarios are given in Table 1 for the upper and lower plena, the core, and the reactor cavity cooling system (RCCS).…”

Section: Phenomena Identification and Ranking Tablesmentioning

confidence: 99%

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

Schultz¹,

Ougouag²,

Nigg³

et al. 2010

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Section: Phenomena Identification and Ranking Tablesmentioning

confidence: 99%

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

Schultz¹,

Ougouag²,

Nigg³

et al. 2010

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“…Possible research topics for the hightemperature helium test section include flow distribution, bypass flow, heat transfer in prototypical prismatic core configurations under forced and natural circulation conditions [9,10], parallel flow laminar instability during pressurized cooldown [11,12], and turbulent heat transfer deterioration [13,14]. Oxidation effects associated with water or air ingress could also be examined [15].…”

Section: Introductionmentioning

confidence: 99%

Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

Sabharwall¹,

O’Brien²,

Yoon³

et al. 2015

EPJ Nuclear Sci. Technol.

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Abstract. A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/ steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750°C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF 4 ) flow loop operating at low pressure (0.2 MPa), at a temperature of ∼450°C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift) in measuring operational data for extended periods of times, as data collected will be used for code and model verification and validation, one of the key purposes for the loop. The experimental facility will provide a much-needed database for successful development of advanced reactors and provide insight into the needs and challenges in instrumentation for advanced high-temperature reactors.

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“…PIRT studies have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Schultz et al, 2006). Consequently, the development of advanced air ingress-related models and verification and validation requirements are high priorities for the NGNP Program.…”

Section: Introductionmentioning

confidence: 99%

Isothermal Air-Ingress Validation Experiments

Kim²

2013

Nuclear Technology

Self Cite

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Idaho National Laboratory carried out air ingress experiments as part of validating comput ational fluid dynamics (CFD) calculations. An isothermal test loop was designed and set to understand the stratified-flow phenomenon, which is important as the initial air flow into the lower plenum of the very high temperature gas cooled reactor (VHTR) when a large break loss-of-coolant accident occurs. The unique flow characteristics were focused on the VHTR air-ingress accident, in particular, the flow visualization of the stratified flow in the inlet pipe to the vessel lower plenum of the General Atomic's Gas Turbine-Modular Helium Reactor (GT-MHR). Brine and sucrose were used as heavy fluids, and water was used to represent a light fluid, which mimics a counter current flow due to the density difference between the stimulant fluids. The density ratios were changed between 0.87 and 0.98. This experiment clearly showed that a stratified flow between simulant fluids was established even for very small density differences. The CFD calculations were compared with experimental data. A grid sensitivity study on CFD models was also performed using the Richardson extrapolation and the grid convergence index method for the numerical accuracy of CFD calculations . As a result, the calculated current speed showed very good agreement with the experimental data, indicating that the current CFD methods are suitable for predicting density gradient stratified flow phenomena in the air-ingress accident.

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Next Generation Nuclear Plant Methods Technical Program Plan

Cited by 15 publications

References 16 publications

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

Isothermal Air-Ingress Validation Experiments

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