RELAP 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-07

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1 Fifth International Seminar on Horizontal Steam Generators 22 March 21, Lappeenranta, Finland. 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-7 József Bánáti Lappeenranta University of Technology Finland David Danielyan Obninsk Institute for Nuclear Power Engineering Russia 1. INTRODUCTION Primary-to-secondary leaks form a significant class of the loss-of-coolant accidents. In a primary-tosecondary leak, the damage is at a place where the containment will be bypassed if the safety valve of the broken steam generator opens. Eventually that can lead to a release of radioactive coolant to the atmosphere. Also, the emergency core-cooling water can escape from the system, which can become a serious problem because it prevents ECC water recirculation. Primary-to-secondary leaks can be divided into three categories; small, medium, and large PRISE. Usually a rupture of a single U-tube (small PRISE) causes the incidents, but the possibility of multiple-tube ruptures (medium PRISE) cannot be ignored completely. The VVER-type reactors have horizontal steam generators where the primary collectors are inside the secondary pool. Construction like that makes also large PRISE's possible. Multiple-tube ruptures and collector breaks may cause the core to be uncovered if sufficient ECC water is unavailable. With certain operator intervention, the flow from the primary to the secondary side may reverse. In such a case, a plug of partially diluted or completely unborated water may be formed on the primary side. If the mixing on the primary side is poor the plug can reach the core and lead to a reactivity accident. Data from several experimental and analytical studies exist for both single- and multiple-tube ruptures. However, the published experiments have focused on the facilities modeling western-type reactors. For the VVER reactors the only published experiment is the SPE on the Hungarian PMK-NVH test facility. The first series of steam generator multiple-tube rupture experiments on PACTEL was carried out in June Those three experiments focused on a possibility of the leak flow reversal occurring as a consequence of operator intervention. The results showed that the reversal is possible under suitable conditions. In the first series, the facility included two steam generators and one accumulator. A new set of experiments was carried out in February 1996 with three steam generators and two accumulators. The main goal was to find the worst possible conditions during which the leak flow reversal can occur. 2. DESCRIPTION OF THE PACTEL FACILITY PACTEL is a volumetrically scaled (1:35) integral test facility modeling the VVER4 reactors used in Finland. The facility includes all the main components of the reference reactor (Fig.1). The reactor vessel is simulated by an U-tube construction consisting of separate core and downcomer sections. The core is comprised of 144 electrically heated fuel rod simulators. The geometry and the pitch of the rods are the same as in the reference reactor. The rods are divided into three roughly triangular-shaped parallel channels, which represent the intersection of the corners of three hexagonal VVER rod bundles. The maximum total core power is 1 MW or 22% of the scaled nominal power. The rods have a nine-step chopped-cosine axial power distribution. The maximum primary pressure is 8. MPa compared to 12.3 MPa of the reference reactor. The component heights and the relative elevations (Fig.2) correspond to those of the full-scale reactor to match the natural circulation pressure heads in the reference system. The hot and cold leg elevations of the power plant have been reproduced. This is particularly important for the correct simulation of loop seal behavior. Unlike other PWRs, there is a loop seal in the hot legs of a VVER4. This is a consequence of the steam generator location, which is almost at the same elevation as the hot leg connection to the upper plenum. The primary collector of the steam generator is connected to the hot leg at the bottom of the steam generator. A roughly U-shaped pipe is needed to complete the connection. The cold leg loop seal is formed by the elevation difference of the inlet and outlet of the reactor 227

2 Figure 1. The PACTEL facility Figure 2. Elevations of the components coolant pump. The number of loops has been reduced from the six of the reference system to three in PACTEL. Thus, one PACTEL steam generator corresponds to two in the power plant. The steam generators have vertical primary collectors and horizontal heat exchanging tubes. The 118 U-shaped heat exchanging tubes are arranged in 14 layers and 9 vertical columns. The average length of the tubes (2.8 m) is about one-third of that in the fullscale steam generator (9. m). The outer diameter of the tubes is 16 mm, which corresponds to the reference system. The inner diameter is 13 mm (in the power plant 13.2 mm). To have a higher tube bundle, the pitch in the vertical direction has been increased to 48 mm instead of the 24mm of the reference steam generator. The pitch in the horizontal direction has been maintained. The outer diameter of the shell is 1. m (in the power plant 3.34 m). Because of the higher vertical pitch, the secondary side is about three times larger than the scaled-down secondary volume. That distorts the time-scale of secondary side transients. Two compartments have been built on each side of the steam generator to decrease the mass of water directly involved in the primary-to-secondary heat transfer process. The compartments are not isolated totally from the rest of the secondary side. The coolant has several flow paths in and out of the compartments. A more detailed description of the facility is in the references. 3. EXPERIMENT PROCEDURE Three new experiments were carried out. The experiments were based on the current instructions for operator intervention during an emergency in the Loviisa nuclear power plant. The first experiment (PSL5) contained primary bleed-and-feed. The bleed-and-feed was realized by opening the pressurizer relief valve and using the HPIS. The second experiment (PSL6) was similar to the first one, but it included a total failure of the HPIS. In the last experiment (PSL7), neither the primary bleed-and-feed nor the HPIS was used. Each experiment had all three steam generators, two intact and one broken (Fig. 3). The main isolation valves were presumed to stick open. That prevented broken loop isolation. 228

3 ORIFICE VALVE Figure 3. The break assembly in the broken SG Table 1. PSL5 PSL6 PSL7 Primary pressure 73.5 bar 75 bar 75 bar Core power 85 kw 85 kw 85 kw Primary mass flow rate 17 kg/s 17 kg/s 17 kg/s Pressurizer heating power 6 kw 6 kw 2 kw Secondary pressure 43 bar 43 bar 43 bar SG feed water mass flow rate 6.5 l/min 6.6 l/min 6.6 l/min SG feed water temperature ~5 o C ~4 o C ~4 o C HPI water temperature ~5 o C ~4 o C ~4 o C Accumulator pressure 53/55 bar 53/54 bar 53/55 bar Accumulator temperature (UP) 17 o C 18 o C 16 o C Accumulator temperature (DC) 18 o C 21 o C 24 o C SIGNAL t = s Lprz < 2.8 m Lprz < 2.3 m Lprz > 7.5 m LSG > 8 cm Lsec < 35 bar Lpri < 54 bar Lprz > 7.5 m ACTIONS Break valve opens Scram Table 2. Pressurizer heaters off, MCPs off, HPI on HPI off - Close MSIV and stop SGFW into broken SG - Open RV and increase SGFW into intact SG - Open pressurizer relief valve Start pressurizer spray Accumulator injection starts Interrupt pressurizer spray 229

4 Table 3. TIME EVENT REASON s Break opens 1 s Core power 15 kw Pressurizer level < 2.8 m 115 s Main circulation pumps stop Pressurizer heaters off Pressurizer level < 2.3 m 355 s Isolation of the broken steam generator Level of the broken steam generator > 8 cm 365 s Intact steam generator relief valve opens Feed of intact steam generators increase 575 s Accumulator injection starts Primary pressure < accumulator pressure 6 s Core power 1 kw 695 s Pressurizer spray on Intact steam generator pressure < 35 bar 134 s Pressurizer spray off Pressurizer level > 7.5 m 39 s eriment ends The experiments focused on finding the worst possible conditions during which a leak flow reversal can occur because of operator intervention. The transients began by the opening of a break (Fig. 3) in a heat-exchanging tube of the broken loop steam generator (steam generator III). The break size (2.5 mm) corresponded to a rupture of five tubes in a reference steam generator. The core power decreased when the collapsed-level in the pressurizer reached 2.8 m. The decay heat curve was followed by reducing the power first to 15 kw and later to 1 kw. When the pressurizer collapsed-level reached 2.3 m, the pressurizer heaters were switched off, and the main circulation pumps (MCPs) started to coast down. In the first experiment, the high-pressure injection (HPI) began from the low pressurizer level signal. Feed water injection into the broken steam generator stopped when the secondary side collapsed-level reached 8 cm. Simultaneously, operator intervention began. The main steam isolation valve (MSIV) of the broken steam generator was closed. The relief valve (RV) in an intact steam generator was opened, and the feed water flow (SGFW) into the intact steam generators was increased. In the PSL5 experiment the pressurizer relief valve was opened. The HPIS was used only in PSL5. The pressurizer spray (1.1 1/min) and the accumulators injected in all the experiments. The initial conditions and operator intervention in the experiments are summarized in Table 1 and 2, respectively. 4. THE EXPERIMENTAL AND CALCULATED RESULTS The general behavior of the experiments was similar at the beginning. When the break was opened in a heatexchanging tube of the broken loop steam generator, the primary pressure and the pressurizer collapsed-level began to decrease. The collapsed-level on the secondary side of the broken steam generator increased. Later, operator intervention and the different boundary conditions made some difference to the behaviour of the system. According to theory and the earlier experience from the old steam generator model, the flow reverses in the lowest tube layers of the steam generators when the facility works in the natural circulation mode. The new steam generators behave same way. The coolant flows in the normal direction at the top of the tube layer and reverses on the bottom. The new set of PRISE experiments shows that it occurs even in the broken steam generator. 4.1 PSL7 experiment Neither the primary bleed-and-feed nor the HPIS was used in this experiment (Fig. 4). As in the experiments PSL-6 and PSL-7, the core power reduced to 15 kw at 1 s. The pressurizer heaters were switched off, and the main circulation pumps began to coast down after 15 s. Then, the steam generation rate on the secondary side decreased, and the secondary side pressure dropped until the pressure control system totally closed the secondary side control valve. The pressurizer was empty at 25 s (Fig. 5). The primary side depressurized until the saturation conditions were reached on the hot side of the facility at 29 s. Then, the primary pressure stabilized at 57 bar (Fig. 6). 23

5 59/ / SPRAY ACCU ACCU / NODALIZATION OF THE PACTEL FOR THE PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENTS HEAT STRUCTURE HEAT SOURCE Figure 4. 5 nodalization scheme of the PACTEL facility VALVES CHECK CLOSING CONTROL SAFETY LEVEL in Presuriser [m] PACTEL PSL-7: Level in the PRZ PRESSURE [bar] PACTEL PSL-7: Pressures in SG (broken) (pri) (pri) (sec) (sec) TEMPERATURE [C] Fig.5. Pressurizer collapsed level PACTEL PSL-7: Temperatures at core outlet 2 Fig.7. Temperature in core inlet. TEMPERATURE [C] 4 35 Fig. 6. Primary and secondary pressures in SG PACTEL PSL-7: Temperatures at core inlet 2 Fig.8. Temperature in core outlet. 231

6 The main steam isolation valve of the broken steam generator was closed at 355 s. After 1 s, the relief valve in the intact steam generator was opened, and the feed water injection into the intact steam generator was increased. The pressure in the intact steam generator dropped, and the primary side cooled. Later, the pressure in the broken steam generator reached the opening pressure of the relief valve. After that, the cycling relief valve controlled the pressure. The accumulator injection started at 575 s. The core power was reduced again 25 s later. These actions effectively dropped the primary temperatures (Figs. 7 and 8). The pressurizer pressure decreased when the pressurizer spray was started at 695 s. To balance the pressure difference between the pressurizer and the rest of the facility, water flowed back into the pressurizer. The hot leg temperatures decreased below the secondary side temperature of the broken steam generator at 88 s. Then, the heat transfer between the primary side of the facility and the broken steam generator secondary side reversed, so the broken steam generator started to act as a heat source. The reversal did not occur in the intact steam generators. The primary pressure balanced with the broken steam generator pressure at 1135 s. Up to this moment, approximately 23 1 (2 1/min) of the primary coolant had flowed into the secondary side. 6 5 PACTEL PSL-7: Mass flow rates in the loop (intact) 6 5 PACTEL PSL-7: Mass flow rates in the loop (broken) MASS FLOW [kg/s] MASS FLOW [kg/s] Fig. 9. Mass flow rate in intact loop 2. Fig. 1. Mass flow rate in broken loop PACTEL PSL-7: Break mass flow rate 15 1 PACTEL PSL-7: Collapsed level in the SG (broken) MASS FLOW [kg/s] COLLAPSED LEVEL [cm] Fig. 11. Break mass flow rate 7 Fig.12. SG collapsed level The pressurizer level was approaching the top of the pressurizer, and the spray system was turned off at 134 s (Fig.5). Then, the flow stagnated in the broken loop, the core outlet temperatures increased, and the decreasing rate of primary pressure slowed significantly. During the rest of the experiment, the flow remained stagnated. By comparing the pressure histories, we see that between 1135 s and 139 s, the pressure difference between the primary and the secondary side of the broken steam generator was big enough for the break flow to reverse (Fig.1). During that time, only about 3 1 (2 1/min) of the coolant on the secondary side of the broken steam generator flowed back to the primary side. The pressures were almost equal during the rest of the experiment. By taking into account the accuracy of the pressure measurement, the pressure difference was too small for a reasonable estimation of the flow rate or 232

7 even a direction of the flow. The secondary side collapsed-level of the broken steam generator did not give any sign of the flow reversal either. The experiment was terminated at 39 s when the primary pressure was about 4 bar. The main events of the experiment are summarized in Table CONCLUSIONS The first series of PACTEL steam generator multiple-tube rupture experiments showed that the leak flow reversal is possible in suitable conditions. Because the first series was made with the old configuration of the facility, a new set of experiments was carried out in February The main goal was to find the worst possible conditions during which the leak flow reversal can occur because of operator intervention. Then, the risk of a partially diluted or completely unborated water plug forming on the primary side exists. The experiments were based on the current instructions for operator intervention during an emergency in the Loviisa nuclear power plant. Leak flow reversal was obvious in the first two experiments where a primary bleed was used. In these two experiments, the average reversed volumetric flow was almost constant. The pressurizer spray and the primary side cooling by the intact steam generators were not enough to reverse the leak flow significantly in the last experiment. The only difference between the first two experiments was the use of the HPIS. By comparing these two experiments, we see that the HPI cools the primary side, but it also slows the rate of decrease of the primary pressure. The accumulator injection has the same effect. The secondary side temperature distribution showed that temperatures are stratified strongly. The coolant was subcooled at the bottom of the steam generator. At the surface, there was a layer of saturated coolant. Hot coolant flowed into the secondary side at the beginning of the transient. Later, the primary coolant was colder than the coolant on the secondary side. However, the secondary side temperature distributions on the broken steam generator did not show any significant mixing of the primary coolant with the original content of the secondary side. The distribution was similar everywhere in the steam generator. So the lower the leakage locates, the longer unborated secondary side water might flow back to the primary side if the flow reverses. The collapsed-level measurements presume the phases are separate. The level measurements of the steam generators showed, however, that the average void fraction on the secondary side was 2 %. That causes errors in the collapsed-level. As it was mentioned earlier the main goal of these calculations was to find out if there would be some reverse flow from secondary to primary side in SG. And the calculations proved that the leak flow reversal is possible in suitable conditions. This shows that the leak flow is significant only when the primary side was actively depressurized. From one hand to minimize the leakage through the SG safety valve, primary pressure must drop as fast as possible. But from the other hand the fast depressurization reverses the leak flow and that way allows a partially diluted or completely unborated water plug to form on the primary side. So, there is a contradiction. Minimizing leakage to the atmosphere and avoiding an ECC water shortage can create conditions that can lead to another accident, which can be really severe. 6. REFERENCES [1] Vesa Riikonen: Steam Generator Multiple-Tube Rupture eriments on the PACTEL Facility with the New Steam Generators. TEKOJA 8/96. December 96. [2] Jari Tuunanen, Heikki Purhonen : General description of the PACTEL test facility. VTT Research Notes. Espoo