Rod Bundle Heat Transfer Test Facility Test Plan and Design (NUREG/CR-6975)
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Manuscript Completed: October 2008
Date Published: July 2010
L.E. Hochreiter, F-B Cheung, T.F. Lin, C. Frepoli, A. Sridharan,
D.R. Todd, E.R. Rosal
The Pennsylvania State University
University Park, PA 16802
K. Tien, NRC Project Manager
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington DC 20555-0001
NRC Job Code W6855
This report describes the program objectives of the Rod Bundle Heat Transfer (RBHT) Program as well as the proposed test design, scaling efforts and the integration of the program into the analysis efforts for improving the best-estimate thermal-hydraulic computer codes. The primary area of investigation is the dispersed flow film boiling processes associated with the reflood portion of a large-break Loss of Coolant Accident. A detailed Phenomena Identification Ranking Table was developed for the reflood process in which the phenomena were subdivided into the individual component phenomena, which a best-estimate computer code models or represents. The individual component models are added in the computer code to provide a prediction of the overall wall heat flux as a function of time during the transient. Since the best-estimate computer codes are modeling individual phenomena on a component level, the experiments and the test instrumentation were developed to provide the detailed information such that the modeling could be confirmed. In this manner, the effects of compensating errors in the modeling will be minimized.
A comprehensive review of other experimental programs has been performed as well as the open literature such that the facility design benefits from the previous experimental work. In addition, a detailed scaling analysis was performed of the facility to determine what, if any, distortion effects could be present which could influence the quality of the experimental data. Both a top down and bottom up scaling analysis was performed using the Zuber-Wulff scaling approach which is state-of-the-art for thermal-hydraulic scaling. The Pi groups were calculated for the facility and for a pressurized water reactor (PWR) and a boiling water reactor (BWR) plant reflood transient and compared. It was found that there is some distortion in the test facility due to material differences of the heater rods relative to nuclear fuel rods, and the radiation heat sink effects of the housing which surround the heater rod bundle. The result was to increase the bundle size, and to investigate the different material types in a separate effects test.
The instrumentation requirements for the facility were driven by the phenomena modeling needs identified in the PIRT. There will be ample instrumentation, as compared to previous tests, to obtain data on void fraction, vapor superheat temperatures in addition to heater rod wall temperatures. In addition, a laser illuminated digital system will be used to measure the entrained droplet size and velocity distributions within the rod bundle. Also, a gamma densitometer may be used to measure the void fraction at fixed locations to compare with the void fraction data from the differential pressure cells. A conceptual design for the test facility has been developed along with a detailed instrumentation plan which addresses the phenomena which was identified in the PIRT. There are over 400 channels of instrumentation for the facility.
The RBHT facility is a unique facility which will provide new data for the fundamental development of best-estimate computer code models. This effort will reduce the uncertainty in the NRC’s thermal-hydraulic computer codes which will enhance the understanding of the complex two-phase phenomena which is modeled for the reflood transient.