Numerical Investigation of Natural Circulation During a Small Break LOCA Scenarios in a PWR-System Using the TRACE v5.0 Code

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1 Numerical Investigation of Natural Circulation During a Small Break LOCA Scenarios in a PWR-System Using the TRACE v5.0 Code ABSTRACT E. Coscarelli, A. Del Nevo University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) Via Diotisalvi 2, 56122, Italy. F. D Auria University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) Via Diotisalvi 2, 56122, Italy. The present paper deals with the preliminary analytical study of the PKL experiment F4.1 performed using the TRACE code (version 5.0). The test F4.1 executed in PKL-III facility investigates both, the boron dilution occurrence and the heat transfer from primary to secondary side under natural circulation conditions (single and two phase flow) and reflux condensation. The boron dilution events in condition of reduced mass inventory, namely during a SB-LOCA scenarios, are considered. The relevance of those transient is connected with the possibility that unborated coolant enters in the core, causing re-criticality or, worse, power excursions. The numerical investigation is performed by developing a complete TRACE input model of the PKL integral test facility, including secondary as well as primary system. The aim of this work is the assessment of the TRACE code against the boron transport and the heat transfer mechanism in the different flow regimes that take place during the experiment. The accuracy of the calculation is evaluated by qualitative and quantitative analysis. The quantification of the accuracy is performed using the Fast Fourier Transform Base Method (FFTBM) developed at University of Pisa. The tool provides an integral representation of the accuracy quantification in the frequency domain. 1 INTRODUCTION The performance assessment and validation of large thermal hydraulic codes and the accuracy evaluation when calculating the safety margins of Light Water Reactors (LWR) are among the objectives of international research programs. In this frame, the conduction of experiments in Integral Test Facilities (ITF), simulating the behavior of Nuclear Power Plants (NPP), plays an important role for the system code assessment and for the identification and characterization of the relevant phenomena during off-normal conditions. OECD set up two test campaigns named SETH projects to be carried out at the PKL test facility, to investigate inadvertent boron dilution events in PWR. Several aspects (simultaneity of natural circulation (NC) restart, minimum boron concentration at RPV inlet, mixing process in SG and RCL, size of the slugs) are relevant and can be analyzed following different initiating events and NPP configurations

2 811.2 The present paper deals with the application and the validation process of the TRACE code, in the framework of code assessment and boron dilution investigation activities carried out at the University of Pisa. The code validation process involves the following aspects: performance of the code, nodalization and user qualification. The user qualification is related to code validation because the user choices affect the code response adopting nodalization solutions and selecting the several available options. The user choices include the tuning of the nodalization parameters to obtain a better response of the code (both in terms of results and resources for the calculation). The above codes have been applied to the simulation of experiments carried out at the PKL facility, within the frame of an international project set up by the OECD which is continuing the PKL III program (detailed information can be found in reference [1]). 2 DESCRIPTION OF THE FACILITY AND EXPERIMENT 2.1 PKL III Test Facility Configuration The PKL facility [2] is a full-height ITF that models the entire primary system (four loops) and most of the secondary system (except for turbine and condenser) of a 1300-MW PWR NPP. Detailed information on the PKL III ITF and data comparison with other PWR test facilities (LOBI, SPES, BETHSY, and LSTF) can be found in reference [3]. The facility includes a Reactor Coolant System (RCS), Steam Generators (SG), the interfacing systems on the primary and secondary side and the break. The RCS includes: The upper head plenum, which is cylindrical, full-scale in height and 1:145 in volume. The upper plenum, full-scale in height and scaled down in volume. The upper head bypass, represented by four lines associated with the respective loops to enable detection of asymmetric flow phenomena in the RCS (e.g., single-loop operation). The reactor core model, consisting of 314 electrically heated fuel rods and 26 control rod guide thimbles. The maximum electrical power of the test bundle is 2512 kw. Thermocouples are located in the rod bundle for measuring the rod temperatures. The reflector gap, located between the rod bundle vessel and the bundle wrapper (the barrel in the real plant). It has a flow resistance designed in order to have 1% of the total primary side mass flow (with MCP in operation) across the reflector gap. The lower plenum, containing the 314 extension tubes connected with the heated rods. The down-comer pipes are welded on the lower plenum bottom in diametrically opposite position. Two plates are located in this zone: the Fuel Assembly Bottom Fitting and the Flow Distribution Plate. The down-comer, modeled as an annulus in the upper region and continues as two stand pipes connected to the lower plenum. This configuration, as already mentioned above, permits symmetrical connection of the 4 CL to the RPV, preserves the frictional pressure losses. The (four) hot legs, designed taking into account the relevance of an accurate simulation of the two phase flow phenomena, in particular CCFL, in the hot leg piping as in the reactor. The (four) cold legs, connecting the SG to the MCP through the loop seal and the MCP to the DC vessel. The hydrostatic elevations of the loop seals are 1:1 compared with the prototype NPP.

3 811.3 The (four) MCPs, which are vertical single-stage centrifugal pumps. The PRZ, full-height and connected through the surge line to the hot leg #2. The SG primary side, modeled with vertical U-tube bundle heat exchangers like in the prototype NPP. The scaling factor has been preserved by reducing the number of tubes (28 tubes with seven different lengths). The SG (secondary side) is constituted by the tube bundle zone, seal welded hollow fillers (below the shortest tubes), the DC (with the upper zone annular containing the FW ring, the central zone modeled by two tubes outside of the SG housing and the lower zone with annular shape) and the uppermost part of the SG that models the steam plenum 2.2 Description of the Test F4.1 RUN 1 The test F4.1 RUN 1 executed in PKL-III facility investigates the boron dilution occurrence in condition of reduced mass inventory (e.g. SB-LOCA event). This is a parametric study of the boron dilution mechanism as a function of the primary mass inventory, such as the already performed test F1.2 but at increased primary pressure (about 3MPa instead 1.2MPa). The boron dilution occurs in the loop seals, at reduced primary mass inventory when the system is operated in reflux condenser mode. The Figure 1 shows the measurement system for detecting the boron concentration during the transient. Two types of measures are available: the continuous measurement of boron concentration performed through COMBO devices described in [3] and the sample measures. In this paper, only measures of the COMBO devices are considered and compared with the code results. Continuous measuring of [B] in Loops #1, 2 and 4 Continuous measuring of [B] in Loop #3 Figure 1 : PKL III experiment F4.1 RUN 1: boron concentration measurement instruments locations.

4 811.4 In the Figures 2 and 3 are shown the trends of the main parameters that characterize the experiment: primary and secondary pressures, primary mass inventory (without PRZ), average mass flow rates, boron concentration in the loop seals, core outlet coolant temperature and cladding temperature in the top of the core. Time (s) 10 4 Time (s) 10 4 (a) (b) Figure 2: (a) Measured trends of loop average mass flow rate and non-dimensional residual mass inventory (b) Measured trends of primary pressure, secondary side pressure, maximum rod surface temperature and core outlet fluid temperature. Boron concentration (ppm) Mass (%) Mass flow (kg/s) Mass (%) Pressare (Pa) Temperature ( C) Time (s) 10 4 Figure 3: Measured trends of boron concentration in loop seal 1 to 4. The results carried out in the natural circulation (NC) experiment F4.1 have been reported in the Natural Circulation Flow Map (NCFM) proposed in reference [3]. The Figure 4 shows that the results are coherent and consistent with the limits established in reference [4] based on calculated transients in ITF. During the test F4.1 RUN 1 the PKL-III facility is operated first in single phase NC conditions and then, following a progressive stepwise reduction of the coolant in the primary side, in two phase flow NC, siphon condensation and reflux condenser modes. Once the core cladding heat up occurs, caused by the low primary mass inventory, the facility is stepwise replenished, with cold water at 2000 ppm of boron concentration, up to the reestablishment of the two phase flow NC conditions in all loops. Boundary conditions, initial conditions and the subdivision of the experiment in different coolant circulation conditions are exhaustively described in references [4]. A brief summary of the main phases is described hereafter in Table 1.

5 811.5 W/P (kg/s/mw) RM/V (kg/m 3 ) Figure 4: Natural circulation maps Table 1: Phenomenological analysis # Phase Sub-Phase Time span [s] Note 0 Start of Test (SoT) Conditioning Phase Reaching of saturated conditions at core outlet / Single Phase natural circulation 2 Single Phase natural circulation During 2 nd drainage 3 Drainage Two phase natural circulation Instability and siphon 4 condensation Reflux-condenser conditions Core uncovered Core dry-out and minimal mass occurrence Reflux-condenser conditions Filling up Instability and siphon condensation / Two phase natural circulation End of Test (EoT) Achievement of the peak mass flow rate between the 4 th and the 5 th drainage 3 PROCEDURE FOR CODE ASSESSMENT The assessment of a TH-SYS code involves the availability of the code, of a qualified nodalization, and of qualified experimental data from a qualified experimental facility. It also requires the availability of a standard procedure fulfilment of specific criteria. In this context, specific references have been carried out at University of Pisa to define the meaning of qualified nodalization and to develop the procedure and the criteria necessary to obtain the qualified nodalization, to perform the code assessment and to execute qualified computer code calculations. For more details see references [5], and. [6]. The code assessment process consists of three main phases hereafter summarized. 1. The steady state results that includes the verifications of the relevant geometrical parameters of the facility (e.g. volume, heat transfer area, elevations, pressure

6 811.6 drops distribution, etc.), which may constitute part of the nodalization qualification process, and the simulation of the nominal steady state conditions against specific acceptability thresholds. 2. The reference calculation results outcome from the qualified nodalization and satisfy qualitative and quantitative accuracy related criteria. It should be noted that the reference calculation is not the best calculation achievable by the code. 3. The results from sensitivity study can be carried out to demonstrate the robustness of the calculation, to characterize the reasons for possible discrepancies between measured and calculated trends that appear in the reference calculation, to optimize code results and user option choices, to improve the knowledge of the code by the user. The attention is focused mainly toward the analysis of the reference calculation results, item 2), considering that steady state calculation, item 1), is part of the nodalization qualification process and sensitivity analyses, item 3), can be designed following the analyses of the item 2). 4 MODELLING OF PKL III BY THE TRACE CODE The TRACE model of PKL facility, shown in Figure 5a, consists of two 3D vessel components in cylindrical geometry that model the rod bundle vessel (RBV) and the RPV down-comer, four separate loops that reproduced the geometry and the hydraulic configuration of the experimental facility. Each one includes a hot leg (HL) a SG, a pump seal, a butterfly valve, a reactor cooling pump and a CL. The pump seal is nodalized with two pipe components: the first one models the circuit from the SG outlet until the BV while the second one reproduces the connection of the loop seal with the RCP. The 3D vessel component that nodalized the reactor core vessel is composed by 45 axial level, two radial rings and six azimuthal sectors. The radial discretization takes in account the internal configuration of the rod bundle vessel characterized by two many radial regions: the reactor core and reflector gap that simulated the side mass flow through the reactor (core bypass). The core bypass hydraulic resistance in the 3D component is introduced using a suitable K- factor so that the mass flow at the bypass during the steady state calculation matched the experimental values that correspond to 1% of the total primary side mass flow. The six azimuthal sector in which is subdivided the cross section of the rod bundle vessel were defined considering the loops arrangement and the down-comer upper head bypass piping disposition (see Figure 5b). The fuel rods in the core region are modeled by six powered fuel rod heat structures, which are arranged in the azimuthal direction each one with the power that corresponds to the sector where there is set out. In the axial direction the fuel rods are nodalized with 18 volumes, the first two level and the last one are not powered, because there represent the unheated length of the core region. The DC vessel model consists of 7 axial levels, 2 radial rings (the inner radius has zero fraction flow area in the radial direction, to reproduce the annular DC model) while the azimuthal nodalization is the same as that of the rod bundle vessel. The down-comer is connected to the RPV by 1-dimensional components that direct the flow from the down-comer to the lower plenum and from the down-comer to the upper head. The four parallel bypass lines that represent the upper heat bypass are model in TRACE with two equivalent parallel bypass pipes. The thermal hydraulic behavior of the pressurizer is simulated through three pipes: the first one, nodalized with one volume, models the bottom of the PRZ, the second one, composed by 20 volumes, analyses the two phase behavior of the pressurizer, finally the last hydraulic component represents the top of the

7 811.7 PRZ, that connects it to the relief steam line, modeled with a pipe. The surge lien is modeled with one dimensional component. The primary side of steam generators are nodalized with a single pipe to represent the U-tube bundle. This hydraulic component, composed by 90 volumes, has the characteristic to preserve the heat exchange flow area between the primary and secondary side. The TRACE model of the steam generator secondary side is composed by three pipes that model respectively the first the rise zone of the steam generator, the second one both the annular top and bottom part but also the two pipes of the down-comer, finally the third one schematizes the dome of the SG. (a) (b) Figure 5: (a) TRACE nodalization of the PKL-2 integral test facility, (b) Azimuthal and radial nodalization of the core region.

8 5 STEADTY STATE CALCULATION The initial conditions of the test have been analyzed and compared using, as a reference, the qualification criteria at steady state level, as envisaged in the UNIPI procedure for code assessment [5]. According with the method the step includes two main tasks: 1. the verification and evaluation of the geometrical fidelity of the model developed, also called the nodalization development phase ; 2. the capability of the analytical model to achieve stable steady state with the correct initial conditions as in the experiment to be simulated. This step is the steady state qualification. The first step, not reported, consists in a systematic comparison between the volumes, the surfaces and the absolute and the relative vertical positions between the system to be modeled and the nodalization. The steady state qualification (step two) deals with the comparisons between the experimental measurements and the calculated results at the Start of Transient (SoT). The selection of the key parameters for the steady state verification was done taking into account the checks requested by the procedure for code assessment [5], and the availability of the experimental data. No error is considered if the calculated value is inside the bands of the measurement accuracy. If it is outside, then, it is calculated as difference between the calculated value and upper or lower limit of the measured value. The stationary conditions of the test are achieved after 1000s of null transient (steady state) calculation in order to stabilize the system. In Table 2 the steady state conditions, calculated by the code, are compared with the experimental value measured at 1000s (from the beginning of the recording) that represent the SoT of the transient. Table 2: Comparison between measured and calculated relevant initial and boundary conditions. # QUANTITY (*) UNIT DESIGN TRACE V5 ER. ACCEPT. ER. ( ), 1 PRIMARY CIRCUIT POWER BALANCE 1-1 Core thermal power MWth % 2.0 % 2 SECONDARY CIRCUIT POWER BALANCE 2-1 SG-1, 2, 3, 4 power exchanged MWth % 3 ABSOLUTE PRESSURE 3-1 PRZ pressure bar % 3-2 Upper plenum pressure bar % 3-3 SG-1 exit pressure bar % 3-4 SG-1 exit pressure bar % 0.1 % 3-5 SG-1 exit pressure bar % 3-6 SG-1 exit pressure bar % 4 FLUID TEMPERATURE 4-1 PRZ fluid temperature C % (**) 4-2 Core inlet temperature (lower plenum top) C % (**) 4-3 Core outlet temperature (upper plenum) C % (**) 4-4 Upper head temperature C % (**) 4-5 HL SG-1 inlet temperature C % (**) 4-6 HL SG-2 inlet temperature C % (**) 4-7 HL SG-3 inlet temperature C % (**) 4-8 HL SG-4 inlet temperature C % (**) 4-9 CL SG-1 outlet temperature C % (**) 4-10 CL SG-2 outlet temperature C % (**) 0.5 % (**) 4-11 CL SG-3 outlet temperature C % (**) 4-12 CL SG-4 outlet temperature C % (**) 4-13 SG feed water temperature (all) C % (**) 5 ROD SURFACE TEMPERATURE 5-1 Clad temperature at 2/3 of the core C C

9 Clad temperature at top of the core C PUMP VELOCITY 6-1 MCP velocity rpm % 10.0 % 7 HEAT LOSSES 7-1 Heat losses primary side kw % 8 LOCAL PRESSURE DROPS 8-1 Pressure drop bar % (^) 9 MASS INVENTORY IN PRIMARY CIRCUIT 9-1 Primary circuit mass inventory (with PRZ) kg % 9-2 PRZ and surge line mass inventory kg % 2.0 % (^^) 10 MASS INVENTORY IN SECONDARY CIRCUIT 10-1 Secondary circuit mass inventory (vessel) from 1 to 4 kg % (^^) 11 FLOW RATES 11-1 CL 1 mass flow rate kg/s % 11-2 CL 2 mass flow rate kg/s % 11-3 CL 3 mass flow rate kg/s % 2.0 % 11-4 CL 4 mass flow rate kg/s % 12 BYPASS MASS FLOW RATES 12-1 Core bypass flow rate (+) % UH-DC bypass flow rate (+) % % 10.0 % 13 PRESSURIZER LEVEL (COLLAPSED) 13-1 PRZ collapsed level m m 0.05 m 14 SECONDARY SIDE OR DOWN-COMER LEVEL 14-1 SGs level (collapsed) m m 0.1 m (^^) 15 BORON CONCENTRATION 15-1 Boron concentration in primary system ppm % -- ( ) The % error is defined as the ratio referenceormesured value calculated value i 100 referenceormesured value The dimensional error is the numerator of the above expression (*) With reference to each of the quantities below, following a one hundred s transient-steady-state calculation, the solution must be be stable with an inherent drift < 1% / 100 s. (**) And consistent with power error. The errors are calculated in K. (^) Of the difference between maximum and minimum pressure in the loop. (^^) And consistent with other errors. (+) This is a design data of the PKL III facility 6 PRELIMINARY POST-TEST ANALYSIS The analysis of the results is based on a comprehensive comparison between measured and calculated trends and values. This is performed in four relevant steps. The first step is based on the qualitative evaluation of the results by making the comparison between values of quantities characterizing the sequence of resulting events. The second step consists of the qualitative evaluation of the results by the comparison between experimental and calculated time trends on the basis of the selected variables. The third step is the qualitative accuracy evaluation of the results on the basis of the Relevant Thermal-hydraulic Aspects (RTA). The qualitative accuracy evaluation is based upon a procedure consisting in the identification of phenomena and of the RTA. The fourth step quantitative accuracy evaluation by using the Fast Fourier Transform Based Method (FFTBM). The intrinsic difficulties of the transient simulation did not allow a complete fulfillment of all the steps in the qualification procedure therefore only a preliminary comparison between the experimental data from the PKL experiment F4.1RUN1 and the results of the TRACE v5 calculations is presented in Figure 6 (only the first 30000s of the transient). Indeed some discrepancies are evident in the comparisons. In particular this difference is evident in first 13000s after the SoT in the temporal evolution of the upper plenum pressure. The cause is mainly related with the boundary conditions imposed for performing the drainage phase. Actually, looking the trend of the primary mass inventory, the mass inventory in the primary system is

10 lower in the code results than in the experiment. This determines an excessive reduction of pressure in the upper plenum zone in comparison with the experimental trend, which is also connected with the not optimized inzilization of the temperature distribution in the secondary side and the set up of the the heat losses in the primary side. After the 13000s from the beginning of the transient the UP pressure behavior is in agreement with the experimental trend. In turn, in Figure 6(a), which reports the the mass flow in the loop 1, it is observed, from the phenomenological point of view, an unstable behavior characterized by oscillations of high amplitude in the time window from 15000s a 30000s. Finally the onset of boron dilution appears slighty anticipated at 26000s due to the difference between calculated and experimental values of the mass inventory. (a) (b) (c) Figure 6: Parameter trends, comparison among experimental data and calculated results (d) 7 CONCLUSIONS This paper illustrates a preliminary analysis, performed by TRACE v5 code, aiming at the evaluation of the code capability the boron dilution occurrence and the heat transfer from primary to secondary side under natural circulation conditions (single and two phase flow) and reflux condensation. The selected experiment is the test F4.1RUN1 belonging to the OECD/CSNI PKL Project. The calculated results of the UP pressure, the mass flow in the loop 1, the boron concentration in the loop seal No. 1 and the primary mass inventory have been reported and compared with the experimental data. Some differences are identified and connected with imperfections in setting up the nodalization and the initial and boundary conditions of the test. In particular, these are the main points which will be addressed in the follow up of the activity: 1. boundary condition for improving the procedure that simulates the drainage form the primary system; 2. initial condition realted to the energy in the steam generator secondary side;

11 set up of the nodalization, namely optimization of the primary circuit heat losses, which influences the energy balance of the system; 4. verification of the pressure drop in the primary system. ACKNOWLEDGMENTS The authors acknowledge the support by the Management board of the OECD PKL Project. REFERENCES [1] K. Umminger, B. Brand, V. Teschendorf, Integral Tests Conducted at the PKL Test Facility on Topical PWR Safety Issues, SETH-PKL3.1, November 5, 2001 [2] Kremin H., et al.; Description of the PKL III Test Facility ; FANP NT31/01/e30, Technical Centre of Framatome ANP Erlangen, Germany, July [3] D Auria F., G. M. Galassi, W. Giannotti, A. Del Nevo, D. Araneo, Assessment of CATHARE2 V2.5 Code Against Boron Transport Experiment, University of Pisa, DIMNP NT 563 (06) Rev. 2, April [4] A. Del Nevo, M. Adorni, F. D Auria, G.M. Galassi, Capability of CATHARE2 V2.5 code in simulating boron dilution following a SB-LOCA. comparison with PKL III results, University of Pisa, DIMNP NT 634 (08) Rev1, January [5] M. Bonuccelli, et al., A Methodology for the Qualification of TH Code Nodalizations, Proc. of NURETH-6 Conference, Grenoble (F), October 5-8, [6] F. D Auria, et al., 2006, State of the Art in Using BE Calculation Tools in Nuclear Technology, Nuclear Engineering and Technology, 38, 2006, pp

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