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Assessment of Critical Heat Flux Correlations in Narrow Rectangular Channels Alberto Ghione, Brigitte Noel, Paolo Vinai, Christophe Demazière

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Assessment of Critical Heat Flux Correlations in Narrow Rectangular Channels Alberto Ghione, Brigitte Noel, Paolo Vinai, Christophe Demazière Assessment of Critical Heat Flux correlations in narrow rectangular channels Alberto Ghione, Brigitte Noel, Paolo Vinai, Christophe Demazière To cite this version: Alberto Ghione, Brigitte Noel, Paolo Vinai, Christophe Demazière. Assessment of Critical Heat Flux correlations in narrow rectangular channels. NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Oct 2016, Gyeongju, South Korea. hal-02124694 HAL Id: hal-02124694 https://hal.archives-ouvertes.fr/hal-02124694 Submitted on 9 May 2019 HAL is a multi-disciplinary open access L’archive ouverte pluridisciplinaire HAL, est archive for the deposit and dissemination of sci- destinée au dépôt et à la diffusion de documents entific research documents, whether they are pub- scientifiques de niveau recherche, publiés ou non, lished or not. The documents may come from émanant des établissements d’enseignement et de teaching and research institutions in France or recherche français ou étrangers, des laboratoires abroad, or from public or private research centers. publics ou privés. NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety Gyeongju, Korea, October 9-13, 2016 . N11P0387 Assessment of Critical Heat Flux correlations in narrow rectangular channels Alberto Ghione (a) , Brigitte Noel Commissariat à l’Énergie Atomique et aux énergies alternatives, CEA DEN/DM2S/STMF/LATF; 17 rue des Martyrs, Grenoble, France [email protected], [email protected] Paolo Vinai, Christophe Demazière (a) Chalmers University of Technology, Division of Subatomic and Plasma Physics Department of Physics; Gothenburg, Sweden [email protected], [email protected] ABSTRACT The aim of the work is to assess different CHF correlations when applied to vertical narrow rectangular channels with upward low-pressure water flow. This is a contribution to the improvement of the thermal-hydraulic modeling of the Jules Horowitz Reactor, which is a research reactor under construction at CEA-Cadarache (France). For this purpose, 46 CHF tests from the SULTAN-JHR experimental database were used. These experiments were performed at CEA-Grenoble in two vertical uniformly heated rectangular channels with gaps of 1.51 (SE3: 20 tests) and 2.16 mm (SE4: 26 tests). The experimental conditions ranged between 0.38 and 0.87 MPa for the outlet pressure, between 1200 and 6600 kg/m 2s for the mass flux, between 56.4 and 156.4 °C for the inlet liquid sub-cooling and between -0.01 and 0.12 for the outlet steam quality. Several models were tested. The Groeneveld look-up tables, which were developed mainly with experiments in pipes, significantly over-estimate the CHF. Furthermore, they fail to predict the decrease of the CHF with the reduction of the gap size. Doerffer’s modification of Groeneveld look-up table for internally heated annuli and the Sudo correlation for nuclear research reactors with plate-type fuel, give better results. In particular, Doerffer’s formula predicts the experimental data with a mean error of -10 % for SE4 and +17 % for SE3, while the Sudo relationship gives mean errors equal to -2.3 % and +32 %. KEYWORDS NARROW RECTANGULAR CHANNELS, CRITICAL HEAT FLUX, SULTAN-JHR, NUCLEAR RESEARCH REACTORS 1. INTRODUCTION Narrow rectangular channels are employed in several engineering systems due to their high cooling capabilities within compact volumes. An example of such an application is the Jules Horowitz Reactor (JHR). The JHR [1] is a material testing reactor under construction at CEA-Cadarache (France). The fuel assemblies consist of cylindrical concentric fuel plates arranged in such a manner that the coolant flows upward through the resulting narrow channels (gap between plates equal to 1.95 mm), experiencing large heat fluxes (up to 5.5 MW/m 2) and high coolant velocities (up to 15 m/s) under nominal conditions. The thermal-hydraulic system code CATHARE [2] is used for the modeling and safety analysis of the reactor. The code relies on a transient 2-fluid 6-equation model, complemented with proper closure laws for single-phase and two-phase flow. These correlations were mainly developed and validated for NUTHOS-11: The 11 th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety Gyeongju, Korea, October 9-13, 2016 . N11P0387 tubes and rod bundles under specific flow conditions typical of commercial reactors [3], thus their applicability to systems with different characteristics has to be carefully scrutinized. For this purpose, the SULTAN-JHR experimental database [4] was employed. The experiments were carried out at CEA-Grenoble during the years 2001 -2008. The test sections consisted of narrow rectangular channels with geometrical parameters (i.e. gap sizes and hydraulic diameters) and system conditions representative of the JHR. The rectangular geometry was chosen to simplify the manufacturing process and to guarantee a high geometric p recision of the test sections. In addition , the curvature of the JHR fuel plates is believed to influence only marginally the flow and the heat transfer [5] [6]. The objective of this paper is to assess the predictive capabilities of selected Critical He at Flux (CHF) correlations against the SULTAN -JHR experimental data. An accurate knowledge of the thermal crisis limits is of crucial importance in the safety analysis of nuclear reactors, since the CHF causes a sharp reduction of the local heat transfer a nd consequently a rapid increase in wall temperatures, leading eventually to burnout. In particular, the 1986 AECL-UO Groeneveld look-up table [ 7] (standard model in CATHARE [8]), the improved 2006 Groeneveld tables [ 9], Doerffer’s formula for internally h eated annuli [10] and Sudo’s correlation [1 1] were tested. The paper is organized as follows: in the next section a brief description of the SULTAN -JHR experiments is given; in Section 3 the correlations selected for the work are summarized along with the ir validity ranges and evaluated against the experimental data; in Section 4 conclusions are drawn. 2. THE SULTAN-JHR EXPERIMENTS AND MODELING In the SULTAN-JHR experimental campaign , about 300 steady-state tests were performed in two narrow vertical rec tangular channel s that were uniformly electrically heated, and where demineralized and degassed water flowed upward. Among these tests, 46 CHF experiments are available . 2.1. Test Section Geometry Two different test sections were used: section 3 (SE3) an d section 4 (SE4) with channel gap equal to 1.509 mm and 2.161 mm, respectively. As shown in Fig. 1, the channel wa s delimited by two Inconel-600 plates, approximat ely 1 mm thick. The power was supplied via direct electrical heating of the plates. The extr emities of the walls are thinner in order to avoid heat concentration effects that may cause early boiling and early CHF at the corners. Fig. 1 Geometry of the SULTAN -JHR test section (top view). The test section wa s encapsulated in an electrical mica -based insulation ( Cogetherm ®) and two pressure steel plates maintained the channel gap and geometry reasonably constant during all tests [12]. The heat losses were significantly reduced with a 200 mm -thick thermal insulation layer made of rock wool. The di mensions of the test section with the associated nomenclature are reported in Table 1. The central part of the channel is heated with an approximately uniform heat flux, while two 70 mm-long adiabatic zones are present at the ends of the test section. A s mooth entrance in the test section was used in order to minimize the entrance effects. NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety Gyeongju, Korea, October 9-13, 2016 . N11P0387 Table 1. Test section geometry (dimensions in mm). Gap size 1.509 2.161 Plate width 47.2 47.15 Length of the corners ( ) 3.15 2.85 Thickness of the corners ( ) 0.5 0.5 Heated height 599.8 599.7 Adiabatic zone height 70.0 70.0 2.2. Instrumentation and CHF detection Several quantities were measured during the experiments, including the mass flow rate, the electrical voltage ∆V and current I, the inlet and outlet water temperatures, the absolute pressures and the pressure drops at several axial locations. The electrical power supplied to the test section could be estimated according to the formula P = ∆V × I . The CHF occurrence is detected with 19 non-isolated thermocouples of type K (called BO-TCs). The thermocouples are located along the width of the channel at a distance of 5 (6 TCs), 15 (7 TCs) and 25 mm (6 TCs) from the end of the heated part of the channel. The sensors are connected to a rapid critical heat flux detection system, which prevents the damage of the test section. The CHF occurs always at the end of the heated channel, since a uniform heat flux distribution is present. A more detailed description of the experimental campaign and facility may be found in [13]. 2.3. CHF tests: procedure and range of conditions To minimize the risk of damage to the test section, a limited number of experiments was performed; in particular, 26 tests in SE4 and 20 tests in SE3. During these tests, the pressure at the exit, the heat flux and the flow-rate were kept constant while the temperature at the entrance was increased by 0.2 °C/min, until CHF was detected. The thermal crisis was considered to start when the BO thermocouples measured a rapid increase of temperature. The parameters at the CHF point were then registered. The range of experimental CHF conditions
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