© 2012 European Nuclear Society Rue Belliard 65 1040 Brussels, Belgium Phone + 32 2 505 30 54 Fax +32 2 502 39 02 E-mail [email protected] Internet www.euronuclear.org ISBN 978-92-95064-14-0 These transactions contain all contributions submitted by 7 December 2012. The content of contributions published in this book reflects solely the opinions of the authors concerned. The European Nuclear Society is not responsible for details published and the accuracy of data presented. 2 of 96 ENC2012-A0026 Development of a thermohydraulic model of the Lazaro Chueca, A. (1); Ammirabile, L. (1); European Sodium Fast Reactor (ESFR) using Martorell, S. (2) the system code TRACE. 1 - JRC-IET, Netherlands 2 - Universidad Politecnica de Valencia, Spain ENC2012-A0028 Generation IV Technology Status including Anderson, G. (1); Lillington, J. (1) recent R & D Activities in ANSWERS 1 - AMEC, United Kingdom ENC2012-A0053 Preliminary Design Assessment of the Molten Merle-Lucotte, E. (1); Allibert, M. (1); Salt Fast Reactor Brovchenko, M. (1); Ghetta, V. (1); Heuer, D. (1); Rubiolo, P. (1); Laureau, A. (1) 1 - LPSC-IN2P3-CNRS / UJF / Grenoble INP, France ENC2012-A0078 Development of materials to withstand the Shepherd, D. (1) extreme, irradiated environments in advanced 1 - National Nuclear Laboratory, United Kingdom nuclear fission reactors ENC2012-A0126 ARCHER:- Material and component challenges Buckthorpe, D. (1) for the Advanced High Temperature Reactor 1 - AMEC, United Kingdom ENC2012-A0258 Pressure Drop Analysis of a Pressure-Tube Type Peiman, W. (1); Saltanov, E. (1); Pioro, I. SuperCritical Water-Cooled Reactor (SCWR) (1); Gabriel, K. (1) 1 - University of Ontario Institute of Technology, Canada ENC2012-A0003 SPES3: THE INTEGRAL FACILITY FOR Ferri, R. (1); Achilli, A. (1); Cattadori, G. SAFETY EXPERIMENTS ON SMALL AND (1); Bianchi, F. (1); Luce, A. (1); Monti, S. MEDIUM SIZED REACTORS (1); Meloni, P. (2); Ricotti, M. E. (3) 1 - SIET S.p.A., Italy 2 - ENEA, Italy 3 - POLITECNICO DI MILANO, Italy ENC2012-A0088 Review of three families of Small Modular Lecomte, M. (1); Beon, J.-Y. (1); Reactors (SMRs): land-based; floating; Poimboeuf, J.-M. (1); Vignon, D. (1) immersed 1 - NucAdvisor, France ENC2012-A0119 PRELIMINARY EVALUATION OF A SEVERE Lo frano, R. (1); Baudanza, V. (1); FLOODING EFFECTS ON AN INNOVATIVE Forasassi, G. (1) SMR. 1 - DIMNP-University of Pisa, Italy ENC2012-A0161 European Design Study on Supercritical Water Schulenberg, T. (1); Starflinger, J. (2); Cooled Reactors Class, A. (1) 1 - Karlsruhe Institute of Technology, Germany 2 - University of Stuttgart, Germany ENC2012-A0202 HEAT-TRANSFER CORRELATIONS FOR Pioro, I. (1); Mokry, S. (1); Gupta, S. (1); SUPERCRITICAL WATER AND CARBON Saltanov, E. (1) DIOXIDE FLOWING IN VERTICAL BARE 1 - University of Ontario Institute of Technology, TUBES Canada ENC2012-A0233 Safety analysis of a sodium-cooled fast reactor Perez-Martin, S. (1); Hering, W. (1); with transmutation capabilities Kruessmann, R. (1); Lemasson, D. (2); Massara, S. (2); Pfrang, W. (1); Ponomarev, A. (1); Schikorr, M. (1); Struwe, D. (1); Verwaerde, D. (2) 1 - Karlsruhe Institute of Technology, Germany 2 - EDF, France 3 of 96 ADVANCED REACTORS 4 of 96 ADVANCED REACTORS I 5 of 96 Development of a thermohydraulic model of the European Sodium Fast Reactor (ESFR) using the system code TRACE. Aurelio Lazaro1,2, Luca Ammirabile1, S. Martorell2 , G. Verdu2 1 - European Commission (EURATOM), Joint Research Centre Institute for Energy and Transport Westerduinweg 3, 1755 LE, Petten, The Netherlands Telf: +31(0)224 54 5446 Email: [email protected] 2- Departamento de Ingeniería Química y Nuclear. Universidad Politécnica de Valencia Cami de Vera, sn, 46021 Valencia. Abstract – One of the main goals of the Generation IV International Forum (GIF) nuclear energy systems is to excel in safety and reliability. To pursue such objective, the development of computational tools able to simulate operation conditions that may be critical for the safety of these innovative reactor concepts is essential. As part of the EURATOM contribution to GIF, the FP-7 CP-ESFR project has been launched to study a Sodium Fast Reactor (SFR) design. This paper presents how a thermohydraulic model of the ESFR plant has been developed using the best-estimate system code TRACE. The model simulates the primary, secondary and tertiary system. The primary system includes the reactor core with point kinetic neutronics implemented and it is able to analyse different accident transient scenarios. The work presented in the paper provides the first steps for the development of the transient part of an integrated safety analysis platform with capabilities to perform detailed simulations of the reactor dynamic thermohydraulic-neutronic coupling techniques. I. INTRODUCTION includes “working horse” designs (pool and loop type, oxide and carbide fuel) that describe cores and systems The Generation IV International Forum (GIF) is an which allow developing and testing different options in a international initiative to develop a new generation of common agreed basis. The design that has been considered nuclear power plants that will excel in safety and reliability in this paper is the pool-type oxide-fuel concept. and will improve in other key-issues as the waste management or the optimization of the fuel usage [1]. This new technology requires specific tools to assess its safety behaviour. To pursue such an objective, the Among the different proposals framed in the GIF, the development and validation of computational tools able to Sodium Cooled Fast Reactor (SFR) has a unique position analyse transients that may affect the plant safety is since related projects have already been developed in essential. several countries for nearly 50 years. A demonstration project within the “European Sustainable Nuclear Industrial In this line, the JRC-IET is developing an integrated Initiative” (ESNII) is planned. The so called ASTRID safety analysis tool [3] with the objective to perform an prototype would be the first Generation IV system based on integrated core and safety analysis of nuclear reactor SFR concept [10]. systems. This platform will assist to fulfil the JRC-IET task to provide independent safety assessment of nuclear In addition to these initiatives, the Collaborative reactors and to contribute in this way to the policy support Project on the European Sodium Fast Reactors, framed in on nuclear safety in the European Union. the 7th Framework Programme, is part of the EURATOM contribution to GIF and merges the efforts of 24 European For its implementation to the ESFR pool-type concept Partners to indentify, organize and implement the R&D it has been pointed as system code the thermo-hydraulic effort needed to develop such a project [2]. This project code TRACE v5.0 [6]. In the following section of this 6 of 96 paper it will be explained how a one-dimensional model of B) and 24 Control and Shutdown Device (CSD) that has been built up and it will be compared with an contains natural boron carbide (19.9% B). equivalent model implemented in the thermo-hydraulic code RELAP5 [5]. A picture of the core layout is shown in Figure 2. II. THE ONE-DIMENSIONAL MODEL Two one-dimensional models have been developed in TRACE and RELAP5 for the ESFR pool-type oxide-core plant following the technical specifications fixed in the Working Horse documents . The plant layout is composed by the pool-type primary system where the heat is generated in the reactor core and transferred via the IHX to the secondary system. Here the heat is extracted from the IHXs and conveyed to the tertiary system where the steam generators produce the steam which drives the turbines, closing the thermodynamic cycle. The secondary system consists of 6 loops, which are thermally linked through the IHX with the pool-type primary and 6 tertiary loops, so that the tertiary system is formed by 36 separate circuits. Fig. 2. Oxide core layout [2]. The pool concept is featured by nearly all the primary The main thermodynamic variables in nominal sodium coolant inside the reactor vessel, enclosing the conditions are listed in Table I. primary pumps and the Intermediate Heat Exchangers (IHX) in addition to the internal structures surrounding the TABLE I core [7]. This configuration is shown in Figure 1. Nominal conditions of the ESFR pool-type plant. Variable Reactor Power (MWth) 3600 Core inlet temperature (˚C) 395 Core outlet temperature (C) 545 Average core structure temperature (˚C) 470 Average fuel temperature (˚C) 1227 II.A. The primary system The thermo-hydraulic configuration of the core is modelled by seven different components according to the core power distribution as proposed in the core technical specifications. These seven components correspond to the hot channel, the inner core zone, the outer core zone (two components), the central dummy assembly and control assemblies, the reflector and the by-pass. Fig. 1. ESFR pool-type concept [2]. These seven core regions (Figure 3) are modelled by a The core is composed by an inner and a outer zone PIPE component, and only six (by-pass excluded) are with different Pu mass content. There are 225 inner fuel attached to a heat structure (HTSTR component) to subassemblies with a Pu mass content of 14.5% and 228 simulate the heat transfer to the coolant. These core outer core sub-assemblies with a mass content of 17%. The components are connected both, to the hot (upper) and cold control rod system is composed of 9 Diverse Shutdown (lower) plena of the plant. The fuel thermal conductivity Devices (DSD) that contains enriched carbon carbide (90% was evaluated according to the Phillipponneau model [9]. 7 of 96 The radial core expansion (diagrid) feedback is calculated based on the expansion of the core support diagrid. The thermal inertia between the sodium core inlet temperature and the average diagrid temperature is taken into account considering a structure of 5 m diameter and 0.05 m thickness.