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2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY Beijing, CHINA, September 22-24, 2004 #Paper C22 COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS T. RADEMER, W. BERNNAT and G. LOHNERT Institute of Nuclear Technology and Energy Systems (IKE), University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany ABSTRACT: The Pebble Bed Reactor plant is a concept of an inherently safe nuclear power plant. This paper presents the coupling of the SIEMENS INTERATOM-developed and licensed 2D neutron diffusion code KIND with the commonly used thermal-hydraulics code THERMIX to simulate very fast transients of High Temperature Reactors (HTR). This is possibly by using more up-to-date thermal-hydraulic models and a more detailed heterogeneous particle model for the fuel kernels. Furthermore, THERMIX is implemented in the steady-state ZIRKUS modular code system which provides necessary initial conditions for the transient code and nuclear data base (macroscopic cross sections) as well. Additionally, a Graphical User Interface (GUI) called SIMPLAT is designed to simplify the input creation and the results analyses for ZIRKUS as well as for KIND. Moreover, the 1D-KIND transient code is coupled with the graphic development software LabView1 to prove the possibility of a runtime interactive intervention of the user during the calculation. This interactive intervention enables to introduce parameter changes while the code is already running. To study the influences of reactivity transients which causes effects that are as fast as the relaxation time for heat exchange between the fuel particles (~ 10 ms) and the surrounding graphite. Results of heterogeneous and homogeneous simulations are compared and show an erroneous overestimation of the maximum power (by one order of magnitude) for the homogeneous model, since the Doppler effect is calculated based on the homogenized and not the real fuel temperature. 0. INTRODUCTION The correct simulation of power distribution, flow and heat transport is of great importance for the next generation high temperature gas-cooled reactors (HTR) such as the “Pebble Bed Modular Reactor (PBMR) (South Africa)” [1] or the HTR-MODUL reactor developed by SIEMENS/Interatom (Germany) [2]. Furthermore it is very important to study the reactor behavior in the case of an accident during the design phase to avoid a failure in design. Future reactors should be easier to build, simpler in their maintenance, inherent safe, and more economical as described by Zanocco in [3]. To achieve this objectives common tools have to be improved to perform the safety related analysis for various designs, e.g. reactivity transients or LOCAs. At the Institute of Nuclear Technology and Energy Systems (IKE), Germany, we are using the kinetic and dynamic neutronic code called RZKIND which is using an own thermal-hydraulic routine called THERM to calculate the temperature fields. This 2-dimensional code was originally developed to simulate the transient behavior of the German pebble bed HTR-MODUL reactor for different accidental scenarios and was licensed by the German licensing authorities (TÜV). This code is also available in a 1D version (ZKIND,WKIND). The code contains a detailed heterogeneous model for 1 Copyrights for “LabView “ by National Instruments -- 1 -- COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS #C22 the heat transport from coated particle fuel to moderator which can simulate very fast reactivity transients much better than a homogeneous model which will be described later on. For the preparation of cross-sections in multigroup approximation the modular program system ZIRKUS developed by Framatome ANP GmbH (former SIEMENS-KWU) was used. It was designed for the realization of various reactor physics calculations for the HTR cores with pebble fuel [4]. It can be used to prepare a cross-section data base for transient calculations by representing the few group cross-sections as a function of different parameters like fuel temperature, moderator temperature, reflector temperature, Xe-density, H2O-density (if water ingress is expected) and control rod positions, which is needed by the transient KIND codes. Additionally, a Graphical User Interface (GUI) called SIMPLAT is in the design phase by our Dept. of Knowledge Engineering and Numerics to simplify the input creation and the results analyses for ZIRKUS as well as for RZKIND. Originally the ZIKRUS input files which are necessary for a calculation had to be prepared by hand. SIMPLAT makes it easier to start and manage the manifold runs of ZIRKUS. Moreover the 1D-ZKIND of the transient KIND code family is coupled with a program development environment called LabView to prove the possibility of a runtime interactive intervention of the user during the calculation. This interactive intervention enables to introduce parameter changes while the code is already running. 1. DESCRIPTION AND CURRENT DESIGN The aim of this work is the development of a simulation package to integrate safety analysis through accident evaluation during early basic engineering stages of the reactor design. One step on this way is the coupling of the thermal-hydraulic code THERMIX/KONVEK with the neutronics code RZKIND and the use of a more detailed heterogeneous particle model for the fuel kernels. The importance for the simulation of very fast transients will be shown. A further step is the implementation of a runtime interactive intervention graphical user interface. 1. 1. Coupling RZKIND with THERMIX/KONVEK The RZKIND code work on the basis of macroscopic one group cross-sections, which are implemented in an associated library as functions of feedback parameters in a polynomial representation. This library is a result of manifold ZIRKUS calculations. Fig. 1 displays the data transformation from ZIRKUS to RZKIND. Many modules have to be used to prepare data for transient calculation with the KIND codes. Like ZIRKUS these modules have data exchange via a general data base by the mass storage archiving and retrieval system MARS [4;6](see Fig.1). RZKIND treats a 2D cylindrical R-Z geometrical structure as well as THERMIX. The KIND code itself consists of three main parts, which are executed sequentially within each time step which is displayed in Fig. 2. Firstly, the neutronic part, to calculate the neutron flux density and the power density in the reactor core using a one group neutron diffusion equation, equations for the delayed neutrons in the usual representation with six groups, equations for the fission product poisoning through Xe-135 (two isotopes) and Sm-149 (chain with six isotopes). Furthermore special treatment for the decay heat (so called pseudo emitters) is implemented. The second part is the thermal-hydraulic part to calculate the temperatures of the fuel elements (solid material) and the coolant gas (helium) in the pebble bed core and in the case of the 2D KIND in the reflector regions. The determination of the helium temperatures in the pebble bed assumes isolated channels with forced streaming. -- 2 -- HTR2004 Beijing, CHINA, 2004. 9. ZIRKUS equilibrium calculations ZIRKUS modular sequences for variations of parameters M A KONTR R 1 group cross sections and boundary S conditions (leakage) for varied parameters D VPF A polynomial expansions T of cross sections A and leakage KINZL, YIKO B DEVA A yields and kinetic parameters S E DATGEN compile RZKIND data file RZKIND data file Figure 1: Flow chart for preparation of RZKIND data base The third part characterizes the neutronic/thermal hydraulic feedback which evaluates modified cross- sections for the neutronic part due to changed parameters. The main parameters are the fuel temperature and the moderator temperatures in the pebble bed, the distribution of the xenon concentration in the fuel elements, the distribution of the water steam concentration in the coolant (accidental water ingress) as well as the position of the control rods and the shutdown elements [4]. The cooling gas passing the pebble bed is treated as flowing in independent radial channels with equal mass flow densities. Only convective heat transfer at the side reflector is considered. This calculation model permits flow rates down to 10% of the nominal mass flow. This restrictions makes it necessary to implement an other thermal-hydraulic tool. Therefore, THERMIX/KONVEK is coupled in a first version with the 2D-RZKIND code, to a coupled version called READY. THERMIX/KONVEK simulates the fluid flow alone with the temperature distribution in the fluid and conducting solids based on a porous media approach which involves the solution of the time dependent energy conservation equation for solids and quasi-steady mass and energy conservation equation together with a simplified momentum equation (Ergun-type with buoyancy) for the gas. For further information see [5] by Nader Ben Said. -- 3 -- COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS #C22 Figure 2: Principle flow chart of the coupled version of THERMIX and KIND dash line the new THERMIX calculation and black line the original THERM path. This coupled version is called READY. In addition there is an appropriate set of empirical correlations for the friction and the pressure drop, modeled by KTA-rule2 3102.3. The properties of the fluid are implemented in accordance to KTA- rule 3102.1. The pressure drop caused by the pebble bed is modeled in accordance to KTA-rule 3102.2 [13; 14; 15]. For all these reasons the THERMIX/KONVEK module is preferred to obtain more accurate solutions; e.g. if natural convection plays a role in a LOFC accident, with flow rates below the limit of 10 % of the nominal mass flow which restricts the THERM routine. Fig. 2 shows the principle sketch of the coupling between the transient code KIND and the thermal- hydraulic code THERMIX/KONVEK. The sketch shows that the former stand-alone code THERMIX is now controlled by the KIND system, but do not replace the original THERM routine; it serves as an alternative, e.g. if natural convection is important in an accidental scenario. Due to the fact that KIND as well as THERMIX have an own time-step control it had been necessary to give one code the leadership. In this case it is KIND which gives the time-step width to THERMIX/KONVEK. 2 KTA rules are Safety Standard from the German Nuclear Safety Standards Commission (KTA) In the internet: http://www.kta-gs.de/ -- 4 -- HTR2004 Beijing, CHINA, 2004. 9. 1600 T-M-0s 1400 KIND-T-M-0s 1200 1000 ] Temperature [ 800 600 400 200 0 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 Height [cm] Figure 3: Temperature distribution over the heights of the HTR reactor model at the beginning of the transient calculation. The black boxes represent temperature distribution of the KIND calculation with the original THERM routine and the red rhombi the results of the calculation with THERMIX 80 70 R Z K IN D (b lu e ) c a lc u la tio n 60 50 Relative Power [-] 40 30 20 P a rtic le M o d e ll o f Z K IN D re d 10 0 0 1 2 3 4 5 6 7 8 9 10 T im e [s ] Figure 4: Relative power for the use of the homogeneous model and the heterogeneous particle model. Calculated with the kinetic codes RZKIND and ZKIND -- 5 -- COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS #C22 Furthermore, the power-density calculated by KIND has to be moved to THERMIX to calculate the temperature field which is used by the KIND code for the next time-step. Fig. 2 shows that the calculation process is divided in two main parts. The stationary branch and the transient branch. Fig. 3 turn out a comparison of the moderator temperature calculated with the THERM module in RZKIND and a calculation with THERMIX, for identical initial conditions after the stationary calculation just at the beginning of the transient calculation. It points out only few differences. The differences in the range of these curves is a result of different net grids used in RZKIND and THERMIX and a different consideration of the extensions in the model. For the future both codes should use the same basic grid net. Further information on recriticality calculations of HTRs can be found in [7; 12]. 1. 2. Coupling of WKIND with TRANSIMOLA To study the possibility of coupling the neutronic code WKIND with a LabView [8] based graphical user interface (GUI), TRANSIMOLA was developed. Principle sketch of the data flow can be seen in Fig. 5. TRANSIMOLA is a user environment to control input data and also the output results. The reasons are to prove if simulations can be performed faster and more effective with an interactive intervention of the user while the calculation is running. The user has faster feedback with the results compared with the normal calculation process. In the normal calculation process starting conditions and the transient behavior like control rod withdrawal or changes of the time-dependent helium inlet temperature as pointwise linear function have to be defined in an input file before the calculation can be started. Then the cross-section data base produced by manifold ZIRKUS calculations is linked to the code and the transient calculation runs. After a problem depending time period the results are written in an ASCII type output file ready to be analyzed. This process requires much time from the definition of a transient problem to the interpretation of the results. Figure 5: left: Principle sketch of the normal calculation process right: Principle sketch of the interactive intervention during simulation In the case of using TRANSIMOLA the starting conditions have to be specified as in the customary calculation process, but the new graphical user interface makes it possible to notice wrong inputs much faster and save time, because the output is directly linked to TRANSIMOLA and gives a real- time feedback to the user. Mistakes can be recognized and calculation runs can be stopped. In addition, TRANSIMOLA allows to change parameters like the control rod position or the helium inlet mass flow rate during the calculation. This allows engineers to change parameters in such a way that the simulation system reacts like a control desk in a reactor and can be used as an educational tool for operators. -- 6 -- HTR2004 Beijing, CHINA, 2004. 9. For these reasons the system is developed with the requirement to run on spatial distributed computers (see Fig. 6). It is possible to start the neutronic code WKIND3, e.g. on a UNIX system and the Graphical user interface on a separate Windows 2000 PC. The data is transferred back and forth over a Data Wheel (Fig. 6), which can be interpreted as an hard disk space which can be reached via a UNIX system and the local PC system of the IKE. Fig. 7 present the user environment of the TRANSIMOLA graphical user interface (GUI), where the input and output variables can be chosen and the transient defined, like for instance the helium inlet temperature, the helium mass flow rate or the control rod position. Fig. 8 and 9 shows sample output screens on TRANSIMOLA. Figure 6: Principle sketch of data flow between 1D KIND and Transimola the LabView user interface. The data is exchanged via the Data Disk an storage file system at the Computer Center Stuttgart. Figure 7: The TRANSIMOLA graphical user interface (GUI) made with the aid of LabView to control the WKIND calculation. You can see the input parameters and the output parameters the user can chose. 3 Beside the UNIX version there exists also a WINDOWS version for PC, WKIND is a 1D Version of KIND -- 7 -- COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS #C22 1. 3. Examples Beside the breakdown of the primary cooling circuit the withdrawal of the control rods is an accident scenario which has to be regarded. In this example a hypothetical extremely improbable reactivity accident is simulated for the HTR-MODUL reactor developed by SIEMENS/Interatom. Fig. 8 and Fig. 9 show the accidental situation in which all the control rods withdraw with a velocity of 1 m/s out of the core up to the end position. There are two SCRAM values implemented simulating the safety systems of the reactor. One checks the relative power/relative flux and the other the helium outlet temperature. If the relative power reaches 1.2 times the nominal value or the helium outlet temperature exceeds a positive temperature increase of 50 Kelvin than a SCRAM signal is given. The nominal thermal power is 200 MW. In our case the SCRAM signal is activated; however , the control rods are not stopped, but get stuck at the end position and will not be inserted - as foreseen. Only the helium mass flow is reduced by controlling the blower (Fig. 8, below). The nominal starting position of the control rods is 240 cm inserted in the active core-zone. The control rods withdrawal with a velocity of 1 m/s out of the core up to the end position by –100 cm. This happens in 3.4 seconds (see Fig. 8 above). The withdrawal of the control rods causes an increase of the reactivity of 1.2 % and effects an increase of the relative power demonstrated in the upper section of Fig. 9. The first scram activates promptly and causes a contrary effect through the decreasing helium mass flow shown in Fig. 8. The result is a peak curve of the relative power. -- 8 -- HTR2004 Beijing, CHINA, 2004. 9. Spos [cm]: control rod position in cm r.m: relative He mass flow rate Figure 8: Fast control rod withdrawal by constant helium inlet temperature and constant helium mass flow rate up to the SCRAM. r.m. is the relative mass flow of helium and Spos is the Control rod position in cm. -- 9 -- COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS #C22 r.P: relative thermal total power Taus [K]: He outlet temperature T,B [K]: fuel temperature Figure 9: In the upper picture the relative thermal power (r.P.) is shown as a result of the fast control rod withdrawal with constant helium inlet temperature and constant He mass flow up to the SCRAM. In the lower picture the helium outlet temperature and the fuel temperature measured in Kelvin. -- 10 -- HTR2004 Beijing, CHINA, 2004. 9. 1. 4. SIMPLAT SIMPLAT is a framework under development by our Dept. of Knowledge Engineering and Numerics to utilize the integration of simulation codes like KIND and ZIRKUS in simulation applications (see Fig. 10). An application consists out of a number of services to allow the user to run a simulation through a certain context. It generates the data sets and control the input generation for all the program modules which are necessary to run a complete transient simulation, archives the outputs and presents the results. By the help of SIMPLAT the work with complex simulations is easier and more error-free and therefore cheaper. Figure 11 shows a typical data sheet of SIMPLAT. Figure 10: Data Flow Chart of the SIMPLAT SYSTEM combined with the ZIRKUS and KIND System -- 11 -- COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS #C22 Figure 11: Typical data sheet of the Info Session of SIMPLAT which contains often changed physical parameters The motivation for this project results in on the claims out of the engineering practice that normally simulation tools could only be handled by specialist. SIMPLAT offers an option to make the knowledge of this experts transparent and for users more understandable. Simulations tools hence become available as a service at the working place. Examples for the outputs in a first version are given in Figure 10 a-d to show that the system is principally working. -- 12 -- HTR2004 Beijing, CHINA, 2004. 9. (a) (b) (c) (d) Figure 12: Examples of different results which are produced by the help of the SIMPLAT system. (a) shows an example form the neutron flux density of energy group 1 (b) the solid temperature in °C over the radius of the cylindrical HTR reactor (c) the decay heat vs time im MW/m³ (d) the power density in the core in MW/m³ 2. CONCLUSIONS AND PROSPECTS The aim of this work is the development of a simulation package to integrate safety analysis through accident evaluation during early basic engineering stages of the reactor design. Therefore the approach of coupling available codes has been to compile a system to simulate neutronic and thermal-hydraulic behavior of the core in the future the complete primary circuit of an HTR inclusive the core itself is envisaged. Stand-alone programs for the transient reactivity calculations and thermal-hydraulics are existing. Research on the capabilities of the codes shows the necessity of improvements. One step has been the coupling of the thermal-hydraulic THERMIX/KONVEK with RZKIND to exempt the neuronic code KIND from the limitations of the THERM routine. An other step has been to use a more accurate heterogeneous model for heat transfer of then fuel kernels to the surrounding graphite for very fast transients, because the homogeneous model overestimates the maximum power by one -- 13 -- COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR REACTORS #C22 order of magnitude, since the Doppler effect is calculated based on the homogenized and not the real fuel temperature. With reference to the paper of A. Walter [10] which couples the CFD code FLOWNEX4 with the KIND the first step to a complete system is done. To simplify the use of this codes the two systems SIMPLAT and TRANSIMOLA are under development. TRANSIMOLA as a graphical runtime interactive intervention tool and SIMPLAT as an development environment and management system for ZIRKUS-KIND calculations. TRANSIMOLA shows in combination with the 1D WKIND code that it can accelerate to obtain solutions through its interactivity by minimizing defective runs and easy output analyses. Changing of cryptic input files which is an error-prone work, can be reduced. Furthermore, the user can intervene during a calculation by changing temporal behavior of the reactor in the simulation, e.g. by changing the control rod velocity or He-mass flow rates. This can be done at any time during the calculation. This ability can be used to simulate simple circuits connected to the core. The neutronic codes are verified by re-calculation of experiments and AVR operating data [12] as well as by comparison with other calculations and reference solutions [7]. Most of the models used in these programs are proven for usual design calculations. In the actual version SIMPLAT manages ZIRKUS and KIND runs; it is still under development. For the future more improvements will be envisaged, e.g. to transfer the 2D KIND into a 3D version and using multi-group data instead of 1-group data used so far. Also a coupling with the CFD code CFX is planned. 4 A thermal-fluid CFD code from M-Tech Industrial (Pty) Ltd., PO Box 19855 Noordbrug 2522, South Africa -- 14 -- HTR2004 Beijing, CHINA, 2004. 9. REFERENCES [1] S. Ion, D. Nicholls, R. Matzie and D. Matzner: Pebble Bed Modular Reactor: The First Generation IV Reactor To be Constructed; World Nuclear Association Annual Symposium 3-5 September 2003 - London [2] G.H. Lohnert: Technical design features and essential safety-related properties of the HTR-MODULE; Nucl. Eng. and Des. 121 (1990) [3] P. Zanocco, M. Giménez and D. Delmastro: Safety design maps: an early evaluation of safety to support reactor design; Nuclear Engineering and Design 225 (2003) 271-283 [4] W. Bernnat, W. Feltes: Models for reactor physics calculations for HTR pebble bed modular reactors; Nuclear Engineering and Design 222 (2003) 331-347 [5] N. Ben Said: The Impact of Design on the Decay Heat Removal Capabilities of a modular Pebble Bed HTR, The 2nd International Topical Meeting on HTR Technology, Beijing, INET China September 22-24, 2004 [6] K. Hoffmann, K.H. Finger, K. Stute: MARS Mass Storage and Retrieval System Documentation, Interatom Nr. 33.05338.4, June 1989 [7] T. Kindt; H. Haque: Recriticality of the HTR-Modul Power Reactor after Hypothetical Accidents; Nucl. Eng. and Des. 137, 107-114 (1992) [8] Labview, User Manual, National Instruments; January 1998 Edition; Part Number 320999B-01 [9] K. H. Finger, K. Hoffmann: MARS: Mass Storage Archieving and Retrieval System Documentation; Interatom; Juli 1990 [10] A. Walter, A. Schultz, G. Lohnert: Comparison of Two Models for a Pebble Bed Modular Reactor Core Coupled to a Brayton Cycle; The 2nd International Topical Meeting on HTR Technology, Beijing, INET China September 22-24, 2004 [11] G.H. Lohnert: The ultimate safety of the HTR-MODULE during hypothetical accidents. In: Proceedings of the Specialists meeting on the decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors, Juelich, Germany. International Atomic Energy Agency, Vienna (Austria) IAEA-TECDOC-757, pp.47-50, 1992 [12] W. Scherer; H. Gerwin, T. Kindt and W. Patscher: Analysis of reactivity and temperature transients at the AVR High-Temperature Reactor; Nucl. Sci. Eng. 97 58- 63 (1987) [13] Nuclear Safety Standards Commission (KTA): Reactor Core Design of High- Temperature Gas-Cooled Reactors: Part 3: Loss of Pressure through Friction in Pebble Bed Cores (March 1981) KTA 3102.3 [14] Nuclear Safety Standards Commission (KTA): Reactor Core Design of High- Temperature Gas-Cooled Reactors: Part 2: Heat Transfer in sperical Fuel Elements (June 1983) KTA 3102.2 [15] Nuclear Safety Standards Commission (KTA): Reactor Core Design of High- Temperature Gas-Cooled Reactors: Part 1: Calculation of Material Properties of Helium (June 1978) KTA 3102.1 -- 15 --

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