atw 2018-04v6

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atw Vol. 63 (2018) | Issue 4 ı April OPERATION AND NEW BUILD 226 European Structural and Investment Funds. This work has been supported by the Project CZ.02.1.01/0.0/0.0/ 15_008/0000293: Sustainable energy (SUSEN) – 2 nd phase realized in the framework of the European Structural and Investment Funds. References [1] CVR Annual Report 2016. [2] http://susen2020.cz/ [3] http://cvrez.cz/en/infrastructure/ research-reactor-lvr-15 [4] IAEA, Standards Safety in the Utilization and Modification of Research Reactors”, Safety Standard n° SSG-24, VIENNA, 2012. [5] ATHLET 3.1A, 2016 User manual: ATHLET Mod 3.1 Cycle a, G. Lerchl, H. Austregesilo, P. Schoffel, D. von der Cron, F. Weyermann, March 2016. [6] Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs), IAEA. http://www-pub.iaea.org/ books/IAEABooks/10731/Heat- Transfer-Behaviour-and-Thermo- hydraulics-Code-Testing-for- Supercritical-Water-Cooled-R [7] TRACE V5.840 Theory Manual, U.S. Nuclear Regulatory Commission, Washington DC, March 2013. [8] TRACE V5.840 User’s Manual, Volume 1: Input Specification, U.S. Nuclear Regulatory Commission, Washington DC, February 2014. [9] TRACE V5.840 User’s Manual, Volume 2: Modelling Guidelines, U.S. Nuclear Regulatory Commission, Washington DC, February 2014. [10] G. Mazzini et al., ATHLET 3.1A SIMULATION CAPABILITIES FOR SUPER- CRITICAL STATE, CVR 1581, 1.1.2017. [11] G. Mazzini et al., ATHLET 3.1A HEAT TRANSFER ASSESMENT FOR SUPER- CRITICAL WATER, CVR 1582, 1.1.2017. [12] G. Mazzini et al., ATHLET 3.1A CAPABILITIES IN SIMULATING SWAMUP FACILITY IN SCW CONDITIONS, CVR 1583, 1.1.2017. [13] M. Polidori, HE-FUS3 Benchmark Specifications, GoFastR-DEL-1.5-01, Rev. 0, ENEA, July 2011. [14] M. Polidori, HE-FUS3 Experimental Campaign for the Assessment of Thermal-Hydraulic Codes: Pre-Test Analysis and Test Specifications, Report RSE/2009/88. [15] M. Polidori et al, HE-FUS3 Benchmark Results, GoFastR-DEL-1.5-6, Rev. 0, November 2012. [16] Miloš Kynčl, Development and Assessment of TRACE HTHL-2 Facility Thermal Hydraulic Model, Internal Project Status Report, CVŘ 1334, March 2017. Authors G. MazziniM. Kyncl Alis Musa M. Ruscak Centrum Vyzkumu Rez (CVŘRez) Hlavní 130 250 68 Husinec – Řež, Czech Republic Numerical Analysis of MYRRHA Interwrapper Flow Experiment at KALLA Abdalla Batta and Andreas G. Class Introduction The MYRRHA reactor, which is developed at SCK-SCN in Belgium, represents a multi-purpose irradiation facility. Its prominent feature is a pool design with the nuclear core submerged in liquid metal lead bismuth. During transients between normal operation and accident conditions decay heat removal is ensured by forced and natural convection, respectively. The flow in the gap between the fuel assemblies plays an important role in limiting maximum temperatures which should not be exceeded to avoid core damage. The term inter-wrapper flow (IWF) describes the convection in the small gap between the wrapper tubes of neighbouring fuel assemblies (FAs). It plays an important role for passive decay heat removal (DHR). Based on numerous experiments several correlations have been proposed for the flow within wirewrapped rod bundles. However, for the flow within the gap between neighbouring bundles only few studies are reported. Recently [1] reviewed the existing correlations by Rheme [2], Baxi & Dalle Donne [3] Cheng and Tordreras [4], and Kirillov [5] for the pressure-drop in wirewrapped rod bundles. The existing correlations were compared to all the available experimental data and showed that agreement of approximately ±20 % can be expected. For the inter-wrapper flow within the gap only few studies exist, see [6]. Due to the scarce database, within the Horizon 2020 – research and innovation framework program of the EU, the SESAME project was established to develop and validate advanced numerical approaches, to achieve a new or extended validation base and to establish best practice guidelines including verification & validation and uncertainty quantification, see [7]. In particular the current work supports the inter-wrapper flow experiment at KALLA. Three fuel assemblies including the gap flow are studied covering the full range of thermo- hydraulic conditions expected in the reactor application. For this purpose, an experimental test matrix has been established which covers relevant scenarios. The aim of our numerical pre-test study is to help the design of the experiment. The current study applied RANS-CFD methods for design support of the experiment. In the body of this compact the experiment, the corresponding numerical model, and preliminary numerical results are provided. 1 Experimental setup The KALLA experiment investigates IWF between three bundles which are thermally connected by a gap. Figure 1 shows a cross-sectional view of the test section which consists of three ducts representing the fuel assemblies. Each duct contains 7 wirewrapped electrically-heated pins representing the fuel rods. The gap between the channels, i.e. assemblies, is filled with liquid metal, so that strong thermal coupling exists between neighbouring assemblies. The test matrix covers independent variation of flow and thermal conditions in both the gap and the bundles. Detailed description of the experiment is reported in [8]. The geometrical parameters of the bundle and the nomenclature are also shown in Figure 1. The experimental loop facility THESYS at KALLA and the Operation and New Build Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
atw Vol. 63 (2018) | Issue 4 ı April | | Fig. 1. SCWL Coolant Maximum Temperature in LOFA. | | Fig. 2. Left: Experimental loop facility THESYS at KALLA showing location where the inter wrapper flow experiment (see Figure 3) will be installed; right: flow diagram for the IWF tests with four parallel channels; the valves V2.1-V2.3 control the flow through the assemblies Q1-Q3. V.2.4 controls the flow in the gap [8]. location where the IWF experiment will be installed is shown in Figure 2 left. Figure 2 right shows the flow diagram of the IWF tests with four parallel channels representing the three assemblies (Q1-Q3) and the gap ( illustrated by the box containing Q1-Q3). The flow and temperature within each assembly and the gap can be set individually by choosing valve openings (V2.1-V2.4) and heating rates according to the KALLA test matrix. Figure 3 shows the geometry of the IWF test section. and mesh resolution for the thermoshydraulic investigation of the gap and the bundle. In particular, we include the upstream components to verify their influence on the flow field within the test section. We employ the k-ε turbulence model and the commercial CFD-code Star CCM+. Our first studied case (i) focuses on the gap | | Fig. 3. Geometry of the IWF test section, dimensions are in mm, the heated part of the bundle is marked red on the left side of the figure, 600 mm, [8]. flow and our second case (ii) on the fuel assembly. For the study of case (i) a computational domain including the lower flow distributer, riser pipe ( including venture tube), upper flow vessel, and the gap are considered (for corresponding technical drawings of components refer to Figure 3). For the study of case (ii) the computational domain includes the lower flow distributer, riser pipe (including venture tube), one inlet expansion and a single 7-pin bundle. Flow properties of the liquid metal Lead-Bismuth eutectic at 200 °C are employed. Note that corresponding upstream pipes and flow conditioners are modelled so that all relevant geometric details are captured. Quantifying the effect of the flow conditioning sections is important for future simulations, as it would enable the use of a simpler computational domain, which still provides accurate results. In the future post-test analysis, the smallest representative computational domain (e.g., potentially without flow conditioner etc.) will be used to compose a fully coupled thermos-hydraulic simulation of the three bundles including the IWF in the gap. Figures 4 left and right show the computational domains for the pre-test studies OPERATION AND NEW BUILD 227 2 Numerical study A comprehensive analysis of the experiment requires efficient simulations. In the pre-test analysis of the hydraulics separate simulations of the gap region and the fuel assembly are performed. In a first step, we determine suitable computational domains | | Fig. 4. Computational domain for IWF-gap (left) and bundle (right) including the upstream domains. Operation and New Build Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
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