Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis.

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Methods are investigated for predicting the power at which critical heat flux (CHF) occurs in TRIGA reactors that rely on natural convection for primary flow. For a representative TRIGA reactor, two sets of functions are created. For the first set, the General Atomics STAT code and the more widely-used RELAP5-3D code are each employed to obtain reactor flow rate as a function of power. For the second set, the Bernath correlation, the 2006 Groeneveld table, the Hall and Mudawar outlet correlation, and each of the four PG-CHF correlations for rod bundles are used to predict the power at which CHF ... continued below

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Feldman, E. & Division, Nuclear Engineering March 30, 2008.

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Methods are investigated for predicting the power at which critical heat flux (CHF) occurs in TRIGA reactors that rely on natural convection for primary flow. For a representative TRIGA reactor, two sets of functions are created. For the first set, the General Atomics STAT code and the more widely-used RELAP5-3D code are each employed to obtain reactor flow rate as a function of power. For the second set, the Bernath correlation, the 2006 Groeneveld table, the Hall and Mudawar outlet correlation, and each of the four PG-CHF correlations for rod bundles are used to predict the power at which CHF occurs as a function of channel flow rate. The two sets of functions are combined to yield predictions of the power at which CHF occurs in the reactor. A combination of the RELAP5-3D code and the 2006 Groeneveld table predicts 67% more CHF power than does a combination of the STAT code and the Bernath correlation. Replacing the 2006 Groeneveld table with the Bernath CHF correlation (while using the RELAP5-3D code flow solution) causes the increase to be 23% instead of 67%. Additional RELAP5-3D flow-versus-power solutions obtained from Reference 1 and presented in Appendix B for four specific TRIGA reactors further demonstrates that the Bernath correlation predicts CHF to occur at considerably lower power levels than does the 2006 Groeneveld table. Because of the lack of measured CHF data in the region of interest to TRIGA reactors, none of the CHF correlations considered can be assumed to provide the definitive CHF power. It is recommended, however, to compare the power levels of the potential limiting rods with the power levels at which the Bernath and 2006 Groeneveld CHF correlations predict CHF to occur.

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  • Report No.: ANL/RERTR/TM-07-01
  • Grant Number: DE-AC02-06CH11357
  • DOI: 10.2172/929269 | External Link
  • Office of Scientific & Technical Information Report Number: 929269
  • Archival Resource Key: ark:/67531/metadc896244

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  • March 30, 2008

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  • Sept. 27, 2016, 1:39 a.m.

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  • Sept. 28, 2016, 6:10 p.m.

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Feldman, E. & Division, Nuclear Engineering. Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis., report, March 30, 2008; United States. (digital.library.unt.edu/ark:/67531/metadc896244/: accessed August 16, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.