%0 Journal Article
%T Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility
%A Jung-Hua Yang
%A Jong-Rong Wang
%A Hao-Tzu Lin
%A Chunkuan Shih
%J Science and Technology of Nuclear Installations
%D 2012
%I Hindawi Publishing Corporation
%R 10.1155/2012/731616
%X In this paper, the TRACE model for IIST facility is developed and verified with the Small Break loss of coolant accident (SBLOCA) experiment of IIST (Institute of Nuclear Energy Research Integral System Test) facility. By using the Wallis and Kutateladze correlations of countercurrent flow limitation (CCFL) model, the TRACE analyses results, such as break flow rate, primary pressure, and the temperature of cold-leg and hot-leg, are consistent with the IIST data. The results show the Kutateladze correlation of CCFL model can well predict the SBLOCA behavior and present good agreement with IIST experiment data in this paper. Besides, the sensitivity study results of Kutateladze correlation in CCFL model are verified and compared with the IIST data. 1. Introduction A reduced-height and reduced-pressure IIST facility has been established for safety studies of the Westinghouse three loops pressurized water reactor (PWR) since 1992. The research purposes of the IIST facility are enhancement of understanding of thermal hydraulics phenomena during the accidents [1¨C3], contributing to evaluate and develop the safety computer codes [4, 5], and validation of EOP during the accidents of PWR [6]. The IIST facility has three loops as well as all the systems associated with Westinghouse PWR plant system transients. The maximum operating pressure of the IIST facility is 2.1£żMPa. CCFL is an important phenomenon in a reactor system. In a PWR, countercurrent flow (CCF) may occur in both the hot-leg and the entrance to the steam generator during LOCA. CCF also occurs during blowdown as the Emergency Core Cooling Systems (ECCSs) fill water into the downcomer. When the CCFL occurs, the mass and heat transfer between gas and liquid phases reduces, and a water pool forms. This phenomenon prevents the core from cooling such that the fuel temperature rapidly increases [7]. Therefore, studies on CCF and CCFL are essential for proper nuclear reactor safety. The codes used in this paper are TRACE v 5.0p2 and SNAP v 2.0.3. TRACE (TRAC/RELAP Advanced Computational Engine) is an advanced and best estimate reactor systems code for analyzing thermal hydraulic behaviors in light water reactors [8]. TRACE consolidates the capabilities of the four codes, TRAC-P, TRAC-B, RELAP 5, and RAMONA, into one modernized code. One of the features of TRACE is its capability to model the reactor vessel with 3D geometry. It can support a more accurate and detailed safety analysis of nuclear power plants. TRACE has a greater simulation capability for loss of coolant accident. Furthermore, a graphic user
%U http://www.hindawi.com/journals/stni/2012/731616/