Title: Comparison of experimental and simulated critical heat flux tests with various cladding alloys: Sensitivity of iron-chromium-aluminum (FeCrAl) to heat transfer coefficients and material properties

Abstract

In this paper we analyze the differences between previous transient critical heat flux (CHF) experiments using iron-chromium-aluminum (FeCrAl), Inconel 600, and stainless steel 316 (SS316) alloy test sections and best-estimate modeling results from widely-used nuclear engineering systems and subchannel analysis tools. FeCrAl is an Accident Tolerant Fuel (ATF) candidate cladding material. The thermal hydraulic performance and safety characteristics of FeCrAl are being evaluated to determine viability as a cladding material in Light Water Reactors (LWRs). In this study, the results of the CHF experiments conducted at atmospheric pressure and fixed inlet coolant temperature and mass flux are compared to models built in the fifth version of the Reactor Excursion and Leak Analysis Program (RELAP5-3D) and CTF, the modernized version of COBRA-TF developed by the Consortium for Advanced Simulation of LWRs (CASL). Results from RELAP5-3D and CTF showed differences from the experiments and from each other in predicting CHF. In the Inconel 600 case, both computational tools overpredicted CHF, which led to an underprediction in the tube outer surface temperature. In the SS316 and FeCrAl cases, CHF was underpredicted by the codes, leading to an overprediction of the tube outer surface temperature. To understand the discrepancies in CHF and post-CHF predictions,more » studies were performed using RELAP5-3D and RAVEN to determine the sensitivity of CHF and peak test section temperature, an analog to peak cladding temperature (PCT), to heat transfer coefficients, a CHF multiplier, and uncertainties in the thermal conductivity and volumetric heat capacity. We found that CHF depends most strongly on the CHF multiplier and thermophysical properties. A combination of these factors that produced the best match to the experiment based on CHF, PCT, and the total energy deposited into the tube was determined. The best match parameters were able to provide best-estimate predictions of the CHF and integral heat flux, but were still conservative when predicting the PCT. The best match set of parameters developed in this paper are intended only as a demonstration of an approach that could be applied in the future with a larger set of experiments to produce more accurate models of CHF and post-CHF behavior.« less

Authors:

Gorton, Jacob P. ^[1]; Lee, Soon K. ^[2]; Lee, Youho ^[2]; Brown, Nicholas R. ^[3]

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+ Show Author Affiliations

1. Univ. of Tennessee, Knoxville, TN (United States); Pennsylvania State Univ., University Park, PA (United States)
2. Univ. of New Mexico, Albuquerque, NM (United States)
3. Univ. of Tennessee, Knoxville, TN (United States)

Publication Date:

Fri Aug 23 00:00:00 EDT 2019

Research Org.:

Univ. of New Mexico, Albuquerque, NM (United States)

Sponsoring Org.:

USDOE Office of Nuclear Energy (NE)

OSTI Identifier:

1848158

Alternate Identifier(s):

OSTI ID: 1558678

Grant/Contract Number:  

NE0008687

Resource Type:

Accepted Manuscript

Journal Name:

Nuclear Engineering and Design

Additional Journal Information:

Journal Volume: 353; Journal Issue: C; Journal ID: ISSN 0029-5493

Publisher:

Elsevier

Country of Publication:

United States

Language:

English

Subject:

73 NUCLEAR PHYSICS AND RADIATION PHYSICS; Nuclear Science & Technology; Accident tolerant cladding; Critical heat flux (CHF); Wettability; Reactivity initiated accidents (RIA); Uncertainty quantification (UQ)