Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

Greenwood, M. S.; Duarte, J. P.; Corradini, M.

doi:10.1016/j.nucengdes.2016.12.030

Title: Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

Full Record
Other Related Research

Authors:: Greenwood, M. S.; Duarte, J. P.; Corradini, M.

Publication Date:: 2017-06-01

Sponsoring Org.:: USDOE

OSTI Identifier:: 1415518

Resource Type:: Publisher's Accepted Manuscript

Journal Name:: Nuclear Engineering and Design

Additional Journal Information:: Journal Name: Nuclear Engineering and Design Journal Volume: 317 Journal Issue: C; Journal ID: ISSN 0029-5493

Publisher:: Elsevier

Country of Publication:: Netherlands

Language:: English

Citation Formats


                    Greenwood, M. S., Duarte, J. P., and Corradini, M. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors.  Netherlands: N. p., 2017. 
Web.  doi:10.1016/j.nucengdes.2016.12.030.

Copy to clipboard


                    Greenwood, M. S., Duarte, J. P., & Corradini, M. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors.  Netherlands.  https://doi.org/10.1016/j.nucengdes.2016.12.030

Copy to clipboard


                    Greenwood, M. S., Duarte, J. P., and Corradini, M. Thu .  
"Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors".  Netherlands.  https://doi.org/10.1016/j.nucengdes.2016.12.030.

Copy to clipboard


                    
@article{osti_1415518,

  title        = {Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors},

  author       = {Greenwood, M. S. and Duarte, J. P. and Corradini, M.},

  abstractNote = {},

  doi          = {10.1016/j.nucengdes.2016.12.030},

  journal      = {Nuclear Engineering and Design},

  number       = C,

  volume       = 317,

  place        = {Netherlands},

  year         = {2017},

  month        = {6}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (Publisher)

Publisher's Version of Record
https://doi.org/10.1016/j.nucengdes.2016.12.030

Other availability

Search WorldCat to find libraries that may hold this journal

Citation Metrics:

Cited by: 4 works

Citation information provided by
Web of Science

Save / Share:

Export Metadata

Save to My Library

Similar Records in DOE PAGES and OSTI.GOV collections:

Experimental investigation of flow transient critical heat flux at light water reactor conditions

Journal Article Lee, Chienhsiung ; Lin, Kuowei

An experimental investigation of flow transient critical heat flux (CHF) for water at pressures 6.9 and 15.5 MPa is conducted under the simulation of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) conditions. The influence of flow decay rates on local mass velocity and quality at CHF is explored to ensure the adequacy of local-condition approach in the CHF predictions. The assessment of transient upstream correction factor to modify the predictions of steady-state CHF correlations during the BWR flow transients are presented. A mechanistic CHF model based on the dryout of a sublayer and CHF correlations are evaluatedmore »« less
Correlation of critical heat flux data for application to boiling water reactor conditions. Final report

Technical Report Hench, J E ; Gillis, J C

A study was conducted of publicly available multirod critical heat flux (CHF) data to generate a correlation which may be used to predict the limits of nucleate boiling in a boiling water reactor. A critical quality-boiling length correlation was generated which includes the effects of rod-to-rod (local) peaking factor, different axial heat flux profiles, pressure, rod diameter and bundle geometry. The CHF data base for this correlation is primarily 16-rod, with heated lengths up to 12 feet, with axially uniform and non-uniform heat flux, in a pressure range from 600 to 1400 psia, and a mass velocity range from 0.25more »« less
In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

Technical Report Upadhyaya, Belle ; Hines, J. Wesley ; Damiano, Brian ; ...

The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. Themore »« less
https://doi.org/10.2172/1183686

Full Text Available
Surrogate materials options for single phase conjugate heat transfer scaled experiments of light water reactors at prototypical conditions

Conference Clement, Simon A. ; Bardet, Philippe M.

In light water reactors, the conjugate heat transfer between heating rods and cooling water is a problem of paramount importance that has been studied for decades. The design of nuclear reactors depends on correlations for the heat transfer coefficient between heat source and coolant, and experimental studies are needed to improve the existing correlations, develop new ones that are adapted to the new reactors designed and their components, and develop high fidelity turbulence models. No studies have been carried out yet at prototypical conditions where the liquid velocity and temperature fields are measured, because of the difficulty to find instrumentsmore »« less
AN EXPERIMENTAL INVESTIGATION OF HIGH-FLUX FREE CONVECTION HEAT TRANSFER TO WATER UP TO NEAR-CRITICAL CONDITIONS

Technical Report Holt, V. E.

An investigation was made to increase the basic knowledge of nucleate and film boiling heat transfer to fluids up to their critical pressures. An extensive literature survey of the subject indicated a scarcity of consistent data and an absence of proven methods for analysis. This situation led to the design and construction of experimental apparatus suitable for measuring rates of free convection heat transfer and associated temperature differences for fluid conditions up to 4000 psi and 800 deg F, with provision for visual observation. The apparatus incorporated a stainless steel system for containing the fluid. Ten-mil-diameter cylindrical and 1/8-in.-high vertical-platemore »« less
https://doi.org/10.2172/4842984

Full Text Available

Similar Records