Uncertainties in modelling and scaling of critical flows and pump model in TRAC-PF1/MOD1

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The USNRC has established a Code Scalability, Applicability and Uncertainty (CSAU) evaluation methodology to quantify the uncertainty in the prediction of safety parameters by the best estimate codes. These codes can then be applied to evaluate the Emergency Core Cooling System (ECCS). The TRAC-PF1/MOD1 version was selected as the first code to undergo the CSAU analysis for LBLOCA applications. It was established through this methodology that break flow and pump models are among the top ranked models in the code affecting the peak clad temperature (PCT) prediction for LBLOCA. The break flow model bias or discrepancy and the uncertainty were ... continued below

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Pages: 22

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Rohatgi, U.S. & Yu, Wen-Shi January 1, 1987.

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Description

The USNRC has established a Code Scalability, Applicability and Uncertainty (CSAU) evaluation methodology to quantify the uncertainty in the prediction of safety parameters by the best estimate codes. These codes can then be applied to evaluate the Emergency Core Cooling System (ECCS). The TRAC-PF1/MOD1 version was selected as the first code to undergo the CSAU analysis for LBLOCA applications. It was established through this methodology that break flow and pump models are among the top ranked models in the code affecting the peak clad temperature (PCT) prediction for LBLOCA. The break flow model bias or discrepancy and the uncertainty were determined by modelling the test section near the break for 12 Marviken tests. It was observed that the TRAC-PF1/MOD1 code consistently underpredicts the break flow rate and that the prediction improved with increasing pipe length (larger L/D). This is true for both subcooled and two-phase critical flows. A pump model was developed from Westinghouse (1/3 scale) data. The data represent the largest available test pump relevant to Westinghouse PWRs. It was then shown through the analysis of CE and CREARE pump data that larger pumps degrade less and also that pumps degrade less at higher pressures. Since the model developed here is based on the 1/3 scale pump and on low pressure data, it is conservative and will overpredict the degradation when applied to PWRs.

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Pages: 22

Notes

NTIS, PC A03/MF A01; 1.

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  • 15. water reactor safety information meeting, Gaithersburg, MD, USA, 26 Oct 1987

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  • Other: DE88006297
  • Report No.: BNL-NUREG-40753
  • Report No.: CONF-8710111-31
  • Grant Number: AC02-76CH00016
  • Office of Scientific & Technical Information Report Number: 5654315
  • Archival Resource Key: ark:/67531/metadc1093214

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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  • January 1, 1987

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  • Feb. 10, 2018, 10:06 p.m.

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  • April 24, 2018, 11:23 a.m.

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Rohatgi, U.S. & Yu, Wen-Shi. Uncertainties in modelling and scaling of critical flows and pump model in TRAC-PF1/MOD1, article, January 1, 1987; Upton, New York. (digital.library.unt.edu/ark:/67531/metadc1093214/: accessed September 20, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.