The PACTEL facility is designed to model the thermal-hydraulic behavior of VVER-440 type pressurized water reactors (PWR) currently used in Finland. These reactors have several unique features that differ from other PWR designs. PACTEL simulates all the major components and systems of the reference PWR, making it a realistic tool to examine a broad range of postulated accidents and operational transients.
PACTEL is a volumetrically scaled (1: 305) facility including a pressurizer, high and low pressure emergency core cooling systems, and accumulators. The maximum operating pressures on the primary and secondary sides are 8 MPa and 4.6 MPa, respectively. The corresponding values in VVER-440 are 12.3 MPa and 4.6 MPa. The reactor vessel is simulated with a U-tube construction including separate downcomer and core sections. The core itself is consists 144 full-length, electrically heated fuel rod simulators with a heated length of 2.42 m. The axial power distribution is a chopped cosine with a peaking factor of 1.4. The maximum total core power output is 1 MW, 20 % of scaled full power. The fuel rod pitch (12.2 mm) and diameter (9.1 mm) are identical to those of the reference reactor. The rods are divided into three roughly triangular-shaped parallel channels representing the intersection of the corners of three hexagonal VVER rod bundles.
Component heights and relative elevations correspond to those of the full scale reactor to match the natural circulation gravitational heads in the reference system. The hot and cold leg elevations of the reference plant have been maintained, including the loop seals. The hot leg loop seals are a result of the steam generator locations, which are only slightly higher than the hot leg connection to the upper plenum. The hot and cold leg connections to the steam generators are at the bottom of each collector, thus a roughly U-shaped pipe is necessary to complete the link to the pressure vessel without sharp bends. The cold leg loop seals are result from by the elevation difference between the inlet and outlet of the reactor coolant pumps, just like in other PWRs.
To preserve flow regime transitions in the horizontal sections of the loop seals under two-phase flow conditions the Froude number has been applied to select the diameter and length of the hot and cold legs. As a result, the total hot and cold leg pipe lengths were shortened by almost a factor of two compared to the reference plant. The pipe cross-sectional area was increased to preserve the volume scaling factor used for the reminder of the facility and preserve the time scale for energy transport from the heat source to the sinks.
Three coolant loops with double capacity steam generators are used to model six loops of the reference power plant. The steam generators have vertical primary collectors and horizontal heat exchanging tubes. The scaled heat transfer area of the tubes is preserved. The 118 U-shape heat exchanging tubes are arranged in 14 layers and 9 vertical columns. The average length of the tubes (2.8 m) is about a third of that in the full scale steam generator (9.0 m). The outer diameter of the tubes is 16 mm,which corresponds to the reference system, and the inner diameter is 13 mm (in the power plant 13.2 mm). In order to have a higher tube bundle the pitch in the vertical direction has been increased to 48 mm instead of the 24 mm of the reference steam generator. The pitch in the horizontal direction has been maintained. Secondary side steam production is vented through control valves directly to the atmosphere. The horizontal orientation of these steam generators is one of the unique features of the VVER design. One consequence of this geometry is a reduced driving head for natural circulation. Another notable feature is the relatively large secondary side water inventory, which tends to slow the progression of transients.
Heat losses for the facility have been measured in a set of natural circulation experiments. Losses from the primary side (excluding the pressurizer) were described in terms of an average core fluid temperature. Primary side heat losses are about 17 kW at 275 °C, dropping to 6 kW at 200 °C, while heat losses from the pressurizer are 3 kW at 265 °C.
The data acquisition system consists of a data acquisition control unit and attached instrumentation. The PACTEL facility is instrumented with temperature, pressure, differential pressure and flow rate transducers. During an experiment about 600 channels are scanned each second. A network of about 400 thermocouples are used to measure temperatures of the primary and secondary side coolant, heater rod cladding, and various structures. Mass flow rates are measured in the downcomer and cold legs using venturi tubes. About 70 differential pressure transducers are used to determine the collapsed water levels in the pressurizer, core, and steam generator secondary side in addition to providing complete circuits of differential pressure measurements around each loop. Test data is written to the data acquisition computer hard disk after each scanning cycle. The modern process control system controls the primary and secondary pressure, core power, and emergency cooling systems.
The PACTEL facility is ideal for investigating planned recovery procedures during accidents and operational transients in VVER-440 plants. PACTEL experiments provide unique data for code developers to validate thermal hydraulic computer codes, such as APROS, RELAP5, and CATHARE, for VVER analyses. The PACTEL facility has provided data for numerous international projects aimed at furthering nuclear power plant safety assessments. The facility is still in operation for example as an auxiliary system for the separate effect test facilities. Until now, 239 experiments have been carried out with the facility.
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