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Preliminary Results of the Seventh Three-Dimensional Aer Dynamic Benchmark Problem Calculation

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Preliminary Results of the Seventh Three-Dimensional Aer Dynamic Benchmark Problem Calculation PRELIMINARY RESULTS OF THE SEVENTH THREE-DIMENSIONAL AER DYNAMIC BENCHMARK PROBLEM CALCULATION. SOLUTION WITH DYN3D AND RELAP5-3D CODES. Marek Beník, Jan Hádek Nuclear Research Institute ež plc, 250 68 ež, Czech Republic E-mail: [email protected] ABSTRACT The paper gives a brief survey of the 7th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at NRI ež. This benchmark was defined at the 20th AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a VVER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The VVER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. Transient calculation was made with the system code RELAP5-3D . The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP th Dukovany was adopted for the 7 AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5-3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. 1. INTRODUCTION The seventh three-dimensional hexagonal dynamic AER benchmark problem was defined as a re-connection of the loop No.1 with its main circulation pump in operation in a VVER- 440/213 plant [1]. The reactor core is in hot full power conditions at the end of its first full cycle. The basic reactor core geometric and material parameters were given. No nuclear cross section data as well as kinetic data were available in the benchmark definition. It means that one’s own best estimate neutronic data have to be used in the calculation. The benchmark participants have to use also their own plant input data decks developed according to the needs of their own NPP system codes. Only main geometric parameters of primary and secondary circuit and main characteristics of control and safety systems were given in order to bring into agreement the input data decks. The calculations of the seventh dynamic AER benchmark problem were performed by using of codes DYN3D [2] and RELAP5-3D [3] at the Nuclear Research Institute (NRI, ÚJV) ež. The reason for NRI ež participation at AER benchmark project is the verification of RELAP5-3D system code solution (especially testing of reactor vessel model) against the results received by other best estimate three-dimensional neutronic codes coupled with NPP system or CFD codes – APROS (VTT Espoo, Finland) [4], DYN3D/ATHLET, ANSYS CFX/DYN3D (HZDR, Germany) [5] and other. 2. RELAP5-3D SYSTEM CODE RELAP5-3D is the latest in the series of RELAP5 codes [3]. The RELAP5 (Reactor Excursion and Leak Analysis Program) series of codes has been developed at the Idaho National Engineering and Environmental Laboratory under the U. S. Nuclear Regulatory Commission, the U. S. Department of Energy, and a consortium of several countries and domestic organizations sponsorship. RELAP5 – at NRI ež has been tested on data from experimental facilities - PMK-NVH (Hungary), PACTEL (Finland), RVS-3 (NRI ež – Czech Republic), and on experimental data from transients and tests performed at NPP Dukovany. The process of RELAP5 testing goes on permanently at NRI ež. The RELAP5-3D version contains several important enhancements over previous versions of the code. The most important feature that distinguishes the RELAP5-3D code from the previous versions is the fully integrated, multidimensional thermal-hydraulic and neutron kinetic modelling capacity. The multi-dimensional component in RELAP5-3D was developed to allow the user to more accurately model the multi-dimensional flow behaviour that can be exhibited in any component or region of a LWR system. Typically this will be the reactor downcomer, lower plenum, core, upper plenum and upper head regions. However, the model is general, and is not restricted to use in the reactor vessel. For example, it can be used for describing of steam generator. The component defines a one, two, or three-dimensional array of volumes and the internal junctions connecting them. The geometry can be either Cartesian (x, y, z) or cylindrical (r, θ, z). The number of volumes in a three-dimensional component is limited to 999 volumes. The multi-dimensional neutron kinetic model in RELAP5-3D is based on the NESTLE code [3], [6] which solves the two or four group neutron diffusion equations in either Cartesian or hexagonal geometry. All thermal-hydraulics is removed from the original NESTLE code in its RELAP5-3D version and instead of it the RELAP5-3D thermal-hydraulic model is used. It means that so called internal coupling between RELAP5-3D thermal-hydraulics and NESTLE kinetic model exists. 3. INPUT MODEL OF NPP WITH VVER-440/213 RELAP5-3D was tested on the input deck of Nuclear Power Plant (NPP) Dukovany with VVER-440/213 reactor (the 1st unit). The input model of NPP Dukovany used for the analyses is a detailed 6-loop model which was modified according to the 7th AER dynamic benchmark specifications. The input deck originates in an “international” 3-loops input deck of NPP Jaslovské Bohunice prepared under IAEA regional program “Safety Assessment of VVER-440/213” [7]. This model has been further enlarged and modified according to the specification of NPP Dukovany and needs for more detailed analyses. The current version of the input model is characteristic with 6-loop configuration, detailed modelling of ECCS systems and detailed modelling of both main steam system and feed water system on the secondary side [8]. 3-D TH reactor pressure vessel model has been validated on available transient data from NPP Dukovany (e.g. asymmetrical loop cooling) [9]. The nodalization scheme of the input model is presented in Fig.1 (reactor vessel) and Fig. 2 (primary circuit). 3.1 Main characteristics of input model Main characteristics of nodalization: ° 6 modelled loops, ° RPV consists of 1 three-dimensional (MULTID) object: - 8 azimuthal sectors, - 4 radial sectors, - 17 axial sectors, - some volumes (related to core and reactor head) are disabled (by setting flow areas = 0), - normal one-dimensional equations are used on each of the coordinate direction (MULTID → Word 7 = 1 on Card CCC0001), ° core (including reflector) is simulated by bundle of 1-D channels (pipes) connected to the multidimensional lower plenum and upper plenum, ° detailed steam generator (SG) nodalization, ° detailed description of steam lines (MSL) and main steam headers (MSH), ° detailed description of feed water (FW) system. Main modelled operational control systems: ° reactor power control system (ARM-5S), ° reactor power limiting system (ROM), ° pressurizer (PRZ) heaters and spraying, ° primary make-up (TK) and let-down (TE), ° turbine control system (TVER), ° steam bypass valves to main condenser (BRU-C). Main modelled safety systems: ° reactor trip signals AZ-1, 2, 3, 4, ° Engineered Safety Features Actuation System (ESFAS) signals, ° all Technological Protection of Steam Generator (TPSG) signals, ° all Local Protection of Steam Generator (LPSG) signals, ° safety valves at PRZ and SGs, ° atmospheric dump valves (BRU-A), ° Emergency Core Cooling System (ECCS), ° High Pressure Injection System (HPIS), ° Low Pressure Injection System (LPIS), ° Accumulators (HA). 3.2 Nodalization of VVER/440-213 core The reactor core configuration is shown in Fig. 3. The core was divided into 12 axial layers (2 layers for unheated lower and upper part, 10 layers for heated core part with total length of 244 cm). The both unheated parts were treated as axial reflectors. The location of control rod groups is shown in Fig. 4. The core was surrounded with a hypothetical radial band of reflector cassettes. 349 fuel assemblies are assigned to 49 core thermal channels (pipes). Reflector cassettes are represented by 8 reflector channels (connected to the 3rd radial sector of MULTID object). Every fuel channel consists of 12 axial volumes (1 non-heated volume – lower reflector, 10 active core volumes, and 1 non-heated volume – upper reflector). The details of reactor core thermal-hydraulic model and its coupling with 3-D neutron kinetic model are shown in Fig. 5. 3.3 Neutron cross sections model RELAP5-3D USER option was selected for the calculations [3]. In this option the user creates his own external subroutine which computes the set of cross sections for a single node given by the material type in the node determined from the composition maps, the region average properties in the node specified in the zone maps for node, and the control fractions and insertion directions for any control rods associated with the node. The user may also specify the value of up to four additional variables in each node that may affect the neutron cross sections (fuel burnup, Xe and Sm concentration etc.). This subroutine is further linked with the RELAP5-3D code. The exchange of parameters between user procedure (subroutine userxs.f) and RELAP5-3D reactor core model can be seen in Fig. 6. 3.4 Macroscopic cross sections library The two-group macroscopic cross section library CSLIBR (see Fig. 6) for homogenized fuel assemblies, control rods and reflectors as well as feedback coefficients and reactor kinetics parameters can be generated from HELIOS neutronic library [10], with using of subprogram DYLIE [11].
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