VALIDATION OF RELAP5/MOD3.3 AGAINST THE PACTEL SBL-50 BENCHMARK TRANSIENT ABSTRACT

Transcription

1 VALIDATION OF RELAP5/MOD3.3 AGAINST THE PACTEL SBL-50 BENCHMARK TRANSIENT J. Bánáti 1 *, V. Riikonen 2 ; V. Kouhia 2, H. Purhonen 2 1 Chalmers University of Technology, SE Gothenburg, Sweden 2 Lappeenranta University of Technology, P.O. Box 20, FI Lappeenranta, Finland joska@nephy.chalmers.se, vesa.riikonen@lut.fi, virpi.kouhia@lut.fi, heikki.purhonen@lut.fi ABSTRACT The PWR PACTEL test facility has recently been designed to support the safety studies of EPR type nuclear reactor thermal-hydraulics. The facility is located at the Lappeenranta University of Technology (LUT) in Finland. It is essentially important to understand the system behavior under natural circulation conditions during a Loss of Coolant Accident (LOCA). With this objective in mind, an international benchmark transient was conducted at the LUT in The SBL-50 test was a SB-LOCA with a 1 mm break in the cold leg. The continuous inventory loss led to core dry-out. This project gave unique opportunities for several organizations to build and validate their models for the PWR PACTEL, as well as to simulate the transient by using various computer codes. Chalmers University of Technology participated with RELAP5/Mod3.3 calculations both in the pre-test and post-test phases of the project. The pre-test simulation included a simplified steam generator model, with the description of the heat exchange by a single characteristic U-tube. This coarse nodalization resulted in reasonably good agreement with the measured data. As the test data became known in the post-test, minor modifications contributed to achievement of better results. The changes were related to the upper plenum nodalization and the critical discharge flow parameters at the break assembly. Even if the post-test model provided better agreement for most of the parameters, it still had difficulties to predict the temperatures in the longest tubes in the steam generators (SGs). A careful examination of the measured data indicated that discrepancies might originate from a flow reversal in the SG primary side. Thus, an advanced SG model was prepared with application of a multi-channel system. Altogether 51 heat exchanger tubes are arranged into 5 bundles in the PWR PACTEL SG, according to 5 different lengths. These bundles were individually modeled in the refined input. The most recent multi-channel SG model has not only confirmed the presence of reverse flow and internal circulation in the SG primary, but it has also contributed to a much better prediction of the fluid temperature distribution. * Corresponding author

2 1. INTRODUCTION Natural circulation can be an effective way of residual heat removal during certain types of transients or off-normal conditions in pressurized water reactors. The mechanism of transporting the decay heat out from the core becomes more enhanced during an accident with gradual decrease of the primary side inventory, such as a loss of coolant accident (LOCA). Several studies have been reported about experimental and analytical investigation of LOCAs with various break sizes. For simulation of the key events and prediction of all the possible consequences on nuclear safety, different types of computer codes have been developed all over the world. However, it is necessary to emphasize that the applicability field of these codes have limitations. Consequently, validations and verifications are always necessary, in order to achieve a qualified model to be used with a qualified code. In addition, users of the same code can form analytical models of the same installation in completely different ways. Also, various codes may produce different results even with using similar analytical approaches. Therefore, it is essentially important for quantification of uncertainty originating from wide spectrum of sources, such as user effects, code effects, or modeling effects. Benchmarking is one the most suitable ways for comparison of different methodologies, gaining experience, and finding best practices with system code calculations. With this objective in mind, several organizations conducted various types of experiments on integral test facilities and separate effect test facilities during the last few decades. Among those, Lappeenranta University of Technology (LUT) was the host for the OECD International Standard Problem, ISP-33 in 1993 [1], which was a stepwise inventory reduction type of experiment conducted in the PACTEL facility, located at LUT in Finland. The facility has been extended and now the PWR PACTEL is used also for studying the safety of the European Pressurized Water Reactor (EPR) being constructed in Olkiluoto, in Finland. The facility is described in [2], [3], and [4] in more details. A Small Break Loss of Coolant Accident (SB LOCA) type experiment, the SBL-50 was offered by LUT for the code user community, as a benchmark exercise. Altogether seven organizations participated in the simulation exercises. The benchmark consisted of two distinct phases: the blind (also known as the pre-test) phase and an open (post-test) phase. A general description, the initial and boundary conditions of the SBL- 50 were introduced at the first workshop, which was held in Lappeenranta on 5 th of October, The participants created their models to be used in the blind phase, basically with knowing only the condition of the termination of the test. Four different system codes were utilized in this phase of the exercise. After sending a well-defined set of the requested parameters, which was a precondition for receiving the measured data from the organizers, the participants could modify their models. The second workshop was held at LUT between 12 th and 13 th of May, This forum was used for discussion of the results of blind and open calculations, share user experiences, as well as modeling difficulties. Summary reports on the results of the blind and open calculations were published in [5] and [6], respectively. The objective of this paper is to summarize the achievements of Chalmers University of Technology in the entire process of modeling the SBL-50 benchmark transient using RELAP5/Mod3.3 computer code. Moreover, the critical and challenging topics of the simulation are presented later in the document.

3 2. THE PWR PACTEL FACILITY The original PACTEL test facility has been designed during the last few years in order to study thermal-hydraulics and safety of the EPR reactors. The PWR PACTEL utilizes some parts of the original facility, such as the core, the reactor pressure vessel, the upper plenum, the downcomer, and the pressurizer [7]. The new PWR PACTEL design is equipped with 2 loops with hot and cold legs, 2 vertical steam generators with hot and cold downcomers and divider plates, and the emergency core cooling systems (ECCS). With this arrangement, the original PACTEL is still maintained alongside with the PWR PACTEL construction. The new structure of the loops and the vertical inverted U-tube steam generators of the EPR style allow experimental research of nuclear thermal-hydraulics to be carried on the field of PWR reactors, particularly for the EPR geometry. The main features of the facility are summarized in Table 1. Table 1 Main characteristics of PWR PACTEL Characteristics PWR PACTEL Reference power plant PWR / EPR Volumetric scale: pressure vessel, steam generators, pressurizer 1:405, 1:400, 1:565 Height scale: pressure vessel, steam generators, pressurizer 1:1, 1:4, 1:1.6 Maximum power [MW] 1 Number of rod simulators 144 Number of primary loops 2 Number of U-tubes in steam generator 51 Average steam generator U-tube length [m] 6.5 Steam generator U-tube diameter / thickness [mm] / 1.24 Maximum primary / secondary pressure [MPa] 8.0 / 5.0 Maximum primary / secondary temperature [ C] 300 / 260 Maximum rod cladding temperature [ C] 800 Main material of components Stainless steel Insulation material Mineral wool The pressure vessel model in PWR PACTEL comprises of an asymmetrical U-shape construction, including the downcomer, lower plenum, core, and upper plenum. The rod bundle of the core consists of altogether 144 (i.e. 3 x 48) electrically-heated fuel rod simulators arranged in three parallel channels. The total electric heating power of the core can be 1 MW. The ECCS of PWR PACTEL includes high and low pressure pumps and two separate accumulators for injection of water to the downcomer and to the upper plenum. The main design focus with PWR PACTEL (Fig. 1) is set on the new construction of the primary loops and vertical steam generators. The set-up is designed to allow simulation of PWR and EPR features and studying loop and steam generator behaviour in particular. Both primary loops simulate one reference EPR loop. The heat transfer area of the steam generator U-tube bundles and the primary side volume of each steam generator scaled down with a ratio of 1/400 compared to the reference steam generator. The inner diameter of the steam generator U-tubes in PWR PACTEL is the same as in the EPR steam generator. The secondary sides of the steam generators include a downcomer, riser and steam dome volumes as well as feed water injection

4 systems. The riser and downcomer parts are also divided into hot and cold parts. There are altogether about 250 temperature, pressure, and differential pressure measurement transducers attached to allow deeper analysis of especially vertical steam generator behaviour. A view of the vertical steam generator facility is presented in Fig. 2 and its characteristics can be found in Table 1. Steam Dome 5 Tube Bundles Hotleg and Coldleg Connections Figure 1 The PWR PACTEL facility Figure 2 Structure of the vertical SG

5 3. THE SBL-50 BENCHMARK EXPERIMENT The SBL-50 was chosen to be the PWR PACTEL benchmark experiment from a series of small break LOCA type test. This experiment was performed earlier to study natural circulation behaviour over a continuous range of primary side coolant inventories. During the blind test phase, only a limited number of parameters were released to the participants. The most important initial and boundary conditions are summarized in Table 2. Table 2 Initial and boundary conditions Parameter Value Primary pressure [MPa] 7.55 (±1) Secondary pressure [MPa] 4.2 (±0.6) Core power [kw] 155 (±6) Pressurizer collapsed level [m] 5.7 (±0.2) Steam generator collapsed levels [m] 3.9 (±0.12) Steam generator 1 / 2 feedwater temperatures [ C] 23 / 19 (±1) Steam generator 1 / 2 feedwater flowrates [dm 3 /min] 1.5 (±0.4) Mass flow in loop 1 / loop 2 [kg/s] 0.60 (±0.14) Quasi steady-state period with closed break [s] Core temperature at the transient termination [ C] 350 Integrated break mass at the end [kg] ~ 300 The SBL-50 experiment was started by a stable operational period at the full. The core power was set to 155 kw as well as the primary and secondary side pressures were initially converging to 75 bar and 42 bar respectively. The actual transient started when the pressurizer heaters were switched off and the pressurizer was isolated. These actions were followed by the opening of the break in the cold leg 2 between the loop seal and the downcomer to produce a slow inventory loss. An orifice plate (Ø1 mm, corresponding to approximately 0.04 % of the PWR PACTEL cold leg cross sectional area) was used to simulate the break. The experiment was terminated when the top of the core dried out and the core temperatures exceeded the value of 350 ºC. No other pre-planned actions were taken during the experiment than maintaining the secondary side water level through an addition of feed water. Yet, the benchmark experiment involved also an unintentional quasi steady-state period that was actually thought to bring more challenge to the benchmark calculations. The break valve closed because of a failure in the valve control at the time when primary system had about 53 % inventory. The incident was discovered and fixed about 20 minutes later. 4. COMPUTER CODE SIMULATIONS The code used throughout the entire calculations presented in this study was RELAP5/Mod3.3hh Patch 3, without any modification in the source. During the development of the current PWR PACTEL model, some parts of an earlier input (prepared for 3 loops and horizontal steam generators at LUT) were re-used. Various input models of the PWR PACTEL were applied in different phases of the benchmark project. Reasons of modifications, their consequences on the transient results, and the steps of model development are summarized in the followings in chronological order.

6 XX XX Upper Plenum SG SG Hot Leg Hot Leg Heat Exchanger Tubes 498 Cold Leg BREAK Cold Leg Downcomer Lower Plenum PRZ Surge Line Core Pressurizer Heat Exchanger Tubes Figure 3 The final RELAP5 nodalization of the PWR PACTEL

7 4.1 The Pre-Test Phase Description of RELAP5 Model The final version of the nodalization scheme of the PWR PACTEL facility is shown in Fig. 3. However, the input model used for pre-test phase was somewhat less detailed: the SG was described with only one characteristic heat exchanger tube (Fig. 4). Lumping of the 51 tubes into one seemed to be adequate since the main goal of the first approach was to keep the model as simple as possible. The core model consists of 3 heated channels (volumes 135, 140, and 145) and one unheated bypass channel (vol. 150). The active section is modeled with 3 separate heat sources of chopped cosine shape. In axial direction, the heating power is distributed into 9 nodes, with addition of 1 unheated volume at the core inlet and 2 volumes at the outlet. The downcomer (vol. 115) is modeled with a straight pipe. The lower plenum (vol. 120) has a U-shape structure consisting of 6 volumes. The exit junction is connected horizontally to the core inlet. The upper plenum (vol. 170) is attached to the core exit (branch 160), where the loops a connected. The facility has 2 basically identical loops, with only one difference: the break assembly (valve 499) is located in loop 2. The hot legs (vols. 200 and 400) are entirely horizontal pipe components. The loop seals (vols. 250 and 450) consist of 12 nodes, each. The PWR PACTEL nodalization includes a pressurizer model (vol. 805). However, this component has a role only in the steady-state 365 To Steam Line m operation period, and it is isolated by the valve 812 at the transient initiation The SG model has a relatively coarse nodalization. The heat exchanger tubes are integrated into a single pipe model. The main concept was to preserve the same average length and the total flow area of the pipes as in the facility. Elevation changes of the adjacent primary and secondary volumes are intentionally set to be identical. This quantity is constant (0.5 m) for the lowest 5 volumes, while it is smaller in the vicinity of the horizontal section of the heat exchanger pipe (vol. 220, red color). The secondary side is divided into the following structural parts: the feedwater is injected to the top of the cold downcomer (vol. 310). The cold side of the riser (vol. 320, blue color) is separated from the hot side (vol. 340, green color) by the divider plate. The hot downcomer is modeled with an annulus component (vol. 350). The boiling section and the steam dome are structurally one unit (vol. 330, yellow color) but with changing flow areas and axial heights. The SG 2 has a similar structure and component numbering scheme as of SG From Hot Leg To Cold Leg Figure 4 Nodalization of SG m m FW Injection m m m m m m m m m

8 4.1.2 Results of the Pre-Test Calculations In order to achieve steady-state conditions with the code, the following strategy was applied. The heaters and spray system of the PRZ were operated within their setpoint pressures, while the core power was kept constant at 155 kw. Since the details of the SG level control system were not released to the participants, a simple proportional-integral type of level controller was designed and applied in the model. This system maintained the SG level constant at its prescribed value (3.9 m) with injecting the necessary amount of feedwater. It was found that the injected flowrate stabilized very close to the specified quantity, and basically all the other parameters converged to a steady condition at 3000 s. The transient run was performed with restarting from the last written block of the steady-state restart-plot file with using the reset option of the code. At time = 0 s, the PRZ was isolated from the system and the break valve was opened simultaneously. The built-in default Henry- Fauske critical discharge flow model was used at the break junction. Based on earlier experiences, the subcooled discharge coefficient was set to 0.8. The following distinctive periods can be recognized during the transient simulation. In the first stage between s the primary system is depressurized, reaching saturation conditions. Between s a slow repressurization can be observed with voiding in the upper plenum. At approx s the collapsed level reached the hot leg elevation and this is reflected in an intensified natural circulation flowrate. Between s the level decrease slowed down in the upper plenum and the two-phase mixture appeared in the hot leg. Steam started to accumulate at the top of the SG tubes. There was a quasi steady-state between 6700 and 7960 s due to the unintentional break valve closure. After reopening of the break, the void penetrated to the core and the heat exchange was in refluxcondenser mode at the SG tubes, while formation of the loop seals began. As Figs. 5 and 6 show, RELAP5 was capable of predicting the key trends of the transient with sufficiently good agreement already in the blind test phase. Matching of the primary system pressure is excellent. Level decrease is somewhat faster in the upper plenum and the core dry-out is delayed compared to the measured values Test P0001 RELAP Test D0062 RELAP Pressure [bar] Level [m] Figure 5 Pressure in the upper plenum 4 Figure 6 Collapsed level in the facility

9 4.2 The Post-Test Phase Modifications in the Model In exchange for the blind test results sent to the organizers, the measured test data was released to the participants. Analysis of the experimental data led to a conclusion that the discrepancies might originated from various sources, such as improper break modeling, rough discretization of the upper plenum, or inadequate distribution of the pressure and heat losses Results of the Post-Test Calculations For better tracking of the level decrease in the upper plenum, the number of nodes has been increased from 4 to 6. Prediction of the heat loss was improved by adjusting the heat transfer coefficient between the facility and the ambient air. However, the most crucial parameter that essentially determines the inventory in the facility is the break flow. Thus, different pairs of critical discharge model parameters were tested for the Henry-Fauske model. It was found that application of C D =0.7 for the discharge coefficient and C Neq =1.5 for the thermal non-equilibrium factor resulted in almost perfect matching with the measured break flow (Fig. 7). The applied changes contributed to a general improvement of the transient simulation (Figs. 8, 9, and 10) Test M9000 CD=0.7 CNeq=1.50 CD=0.8 CNeq=0.14 CD=1.0 CNeq= Test D0062 RELAP5 Mass [kg] Level [m] Figure 7 Integrated break flows 4 Figure 8 Collapsed level Test F0022 RELAP5 300 Test T0116 RELAP Mass Flowrate [kg/s] Pressure [bar] Figure 9 Mass flowrate in cold leg Figure 10 Core exit temperature

10 4.3 Nodalization Study with the Multi-Channel SG Model Modifications of the parameters during the posttest phase had clearly positive effects. This was especially true for the temperatures in the facility. However, despite all efforts, the applied single tube SG model still underpredicted some of the primary side fluid temperatures in SGs. A closer examination of the measured radial temperature distribution revealed that there was an approximately 8 o C difference between the shorter and longer tubes before 3000 s (Fig. 11). One possible explanation for such a behavior is an assumption of internal circulation. According to the test data, there is almost a clear evidence for a forward (hotter) flow in the shorter tube bundle, and a reverse flow (carrying colder fluid from the cold collector) in the longer tubes during onephase natural circulation. Non-uniform flow distribution in SG tubes was observed in other test facilities. Kukita et al [8] and Jeong et al [9] have published studies in the same topic. When the collapsed level reached the hot leg elevation, the natural circulation intensified, and the steam bubbles could penetrate into the SG tubes. At this point, a sudden temperature increase can be observed in the longer tube, while the temperature dropped in the short tube simultaneously. Higher mixing of the flows is indicated by the nearly balanced temperatures after ~3500 s. Later into the transient, the SG is operated in reflux condensation (or boilercondenser) mode. Detailed knowledge of this phenomenon is very important because it can have severe consequences on nuclear safety: e.g. the reverse flow of un-borated water may enter the core and cause re-criticality. In order to capture the uneven distribution of the flow, the single-channel SG model had to be replaced with a multi-channel version (Fig. 12). For this purpose, 5 tube bundles were individually modeled according to their different lengths. Furthermore, the axial discretization was refined by reduction of elevation changes. Temperature [C] SG1 Temperatures (Hot Side, Elev. 700 mm) Short tube No. 10, T1095 Long tube No. 50, T Figure 11 Measured temperatures in the short and long tubes at 700 mm elevation From Hot Leg To Steam Line To Cold Leg Figure 12 Nodalization of the multi-channel SG model FW Injection

11 4.3.1 Results of the Nodalization Study With application of the refined multi-channel SG model, it has been proven that unequal distribution of the flow may exist and it can be simulated by the code (Fig. 13). Distinct behaviour of the longest and the shortest tubes is clearly visible. The first period can be characterized by nearly stable one-phase natural circulation until ~2900 s. There is a reverse flow in the longest tube indicated by constant negative mass flowrate of approx kg/s, while the fluid is flowing in forward direction in the shortest tube. Emptying of the upper plenum intensified the forward flow, and when it reached a peak, the reverse flow started to stagnate. Only forward flow existed with decreasing flowrates after 3800 s. Oscillatory nature of the flow with positive and negative peaks is a typical feature of boiler-condenser mode of heat transfer. Since the SG tubes were not equipped with flow meters, there was no data available for comparison. Considering the temperatures, the single-channel SG model showed the largest discrepancies in comparisons with the longest tube data (Fig. 14). It can be seen that RELAP5 overestimated this parameter by more than 8 o C during the initial period of ~3000 s. The multi-channel model has eliminated this error almost entirely. Matching of the fluid flow temperatures is excellent. The code had only some minor difficulties during the transition period around 3000 s but the subsequent period was again well simulated SG1 Mass Flowrates (Hot Side, Elev. 700 mm) Longest Tube Shortest Tube SG1 Temperatures (Tube 50, Hot Side, Elev. 700 mm) Test T1115 Single-Tube Model Multi-Tube Model Mass Flowrate [kg/s] Temperature [C] Figure 13 Forward and reverse flows in the shortest and longest tubes of SG 1 model 252 Figure 14 Comparison of temperatures in single and multi-tube models of SG 1 5. SUMMARY AND CONCLUSIONS The developmental process of a RELAP5 model was demonstrated for simulation of a SB LOCA-type experiment conducted in the PWR PACTEL facility. The input development had basically 3 stages with certain modifications in between. Sufficiently good results had already been achieved with a simpler description of the SGs. However, the single-channel model was unable to predict a flow reversal. The most detailed multi-channel model has confirmed the presence of reverse flow and it has also contributed to a higher degree of accuracy in prediction of the fluid temperature distribution.

12 Modeling of the break flow is always a challenging task with system codes. Default model parameters or recommended values obtained from user guidelines do not necessarily result in the most precise coolant inventories. However, it has been shown that with variation of the breakrelated coefficients, the integrated break mass flow can be approximated correctly. Importance of the flow resistances, pressure and temperature losses cannot be neglected. The role of these parameters is even more enhanced in the case of such a relatively small facility as the PWR PACTEL, and particularly, during a transient with low power natural circulation. Minor inaccuracies in estimation of the resistances may result in large deviations of flow distributions. 6. REFERENCES [1] Purhonen, H., Kouhia, J., and Holmström, H., OECD/NEA/CSNI International Standard Problem No. 33 (ISP 33) PACTEL Natural Circulation Stepwise Coolant Inventory Reduction Experiment, Comparison Report, Vol. 1-2, NEA/CSNI/R(94)24 Part I-II, [2] Riikonen, V., et al., General Description of the PWR PACTEL Test Facility. Research Report, Lappeenranta University of Technology, Nuclear Safety Research Unit, YTY 1/2009. Lappeenranta, [3] Rantakaulio, A., et al., A New Integral Facility PWR PACTEL for Vertical Steam Generator Simulation. Proceedings of International Congress on Advances in Nuclear Power Plants (ICAPP '10), , San Diego, California, USA ISBN [4] Kouhia, V., et al., PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications. Integral Test Facilities and Thermal-Hydraulic System Codes in Nuclear Safety Analysis, Science and Technology of Nuclear Installations, vol. 2012, Article ID , 8 pages, doi: /2012/ [5] Riikonen, V., Kouhia, V., Summary Report of PWR PACTEL Benchmark Experiment Blind Calculations. Research Report, Lappeenranta University of Technology, LUT Energy, Nuclear Safety Research Unit, PAX 1/2011. Lappeenranta, [6] Riikonen, V. and Kouhia, V., Summary Report of PWR PACTEL Benchmark Experiment Open Calculations. Research Report, Lappeenranta University of Technology, LUT Energy, Nuclear Safety Research Unit, PAX 2/2011. Lappeenranta, [7] Tuunanen, J., et al., General Description of the PACTEL Test Facility. VTT Research Notes, p. + app. 75 p. VTT Technical Research Centre of Finland, Espoo, ISBN [8] Kukita, Y., Nakamura, H., and Tasaka, K., 1988, Nonuniform Steam Generator U-tube Flow Distribution During Natural Circulation Tests in ROSA-IV Large Scale Test Facility, Nucl. Sci. Eng., 99 (1988) pp [9] Jeong, J., et al., Non-Uniform Flow Distribution in the Steam Generator U-tubes of a Pressurized Water Reactor Plant During Single and Two-Phase Natural Circulations, Nuclear Engineering and Design 231 (2004) pp