Experimental studies of the fluid dynamics during a pressurized thermal shock (PTS)
For securing core cooling in case of a loss-of-coolant accident in a nuclear reactor additional emergency cooling water is fed from different storage vessels into the main pipes of the primary circuit that is under pressure. Those main pipes connect the reactor pressure vessel with the steam generator (the so called hot legs, where the heated water flows away from the reactor) and the main coolant pump (the so called cold legs, where the colder water flows towards the reactor). A safety issue is thermo-mechanical stresses introduced to the reactor pressure vessel (RPV) wall by sudden contact with the cold liquid. After reactor shut-down the RPV and the connecting pipes with saturated steam and water inside still have temperatures well above 200°C while the emergency core cooling (ECC) water may be below 50°C. If this ECC water was flowing badly mixed into the RPV downcomer, the vessel wall would be exposed to severe thermo-mechanical loads. The sudden cooling of the wall under pressure (also referred to as pressurized thermal shock) could potentially lead to cracking. The risk of an RPV failure both depends on the actual structural mechanical properties of the wall material as well as on the thermal hydraulic phenomena governing the fluid mixing in the main coolant pipe.
|PTS scenario: In case of a loss-of-coolant accident cold emergency core cooling water will be fed into the hot main coolant pipe. If insufficiently mixed with the hot water in the pipe, the reactor pressure vessel wall may be exposed to high thermal stresses.|
The latter question must be answered by fluid mechanical investigations. The flow dynamics in an ECC scenario is governed by diverse thermal fluid dynamic phenomena, such as thermal mixing of hot and cold water, heat transfer by steam condensation at the ECC jet and the free surface area. Here shape and turbulence of the entering ECC water jet, the amount and intensity of steam bubble entrainment by the jet as well as the liquid turbulence in the main coolant pipe play a major role. The interplay of flow and heat transfer depends in a complex way on the geometric and thermal hydraulic boundary conditions, like ECC mass flow rate and temperature, filling level and temperature of the water in the main coolant pipe as well as the system pressure, which determines the steam density.
|Scheme of the PTS test facility (left) und photo of the TOPFLOW pressure tank (right)|
Within the framework of a consortium project that has been running from 2006-2012 unique investigations have been carried out at the TOPFLOW facility at HZDR. Within the consortium that led the investigations, partners were the energy enterprises EDF France and AREVA NP France, the publicly-funded institute IRSN (National Institute for Radiation Protection and Nuclear Safety) and CEA France as well as Paul Scherrer Institute (PSI) and ETH Zürich from Switzerland. For conducting experiments with flows of steam and water close to reality, the so-called pressure tank at the TOPFLOW facility of the HZDR was used. The experimental set-up models the principal hydraulic components of a French pressurized water reactor to a scale of 1:2.5. It comprises a section of the cold leg with a main pipe including the pump casing, the downcomer section of the reactor pressure vessel as well as the pipe for the emergency coolant. The test rig is superiorly equipped with measuring instrumention, which delivers spatial and temporal high-resolution data, e.g. about the temperature distribution and the flow status in the experimental setup.
|High-speed video image of the ECC jet at different mass flow rates (left). Infrared camera image of the temperature distribution on the cold leg wall showing the impinging cold jet (blue) and the separated steam (red) and water (yellow) in the cold leg (center). CFD simulation, showing the wall temperature distribution between the injection position and the downcomer (right).|
At over 200 measuring points temperatures are recorded in the main pipe as well as in the downcomer section of the reactor pressure vessel, an infrared camera measures the exact temperature distribution of the pipe walls and finally a high-speed camera observes the flow of the emergency water as it enters the main pipe. Apart from that, wire-mesh sensors, which have been developed at the HZDR, are implemented to record the water speed in the main coolant pipe.
The goal of the experiments was to provide validation data for CFD code development. Such developments are for instance carried out in the European projects NURESIM, NURISP and NURESAFE. The experimental series covered more than 90 experiments incorporating different operating parameters in each experiment, i.e. pressure, flow level in the main coolant pipe and mass flow as well as temperature of the ECC water. These variations are imperative to fulfill the requirements of the most important thermo-hydraulic characteristic numbers that can be used to scale up individual effects to plant dimension, and on the other hand to cover a wide range of parameters for current and future CFD simulations. Eventually, comparative experiments with air-water and steam-water were made to assess the effect of steam condensation that is lacking in the air-water experiments.
- EDF, France
- Commissariat à l'énergie atomique et aux énergies alternatives (CEA), France
- Institut de Radioprotection et de Sûreté Nucléaire (IRSN), France
- AREVA NP, France
- Paul-Scherer-Institut (PSI), Switzerland
- Eidgenössische Technische Hochschule Zürich (ETH), Switzerland
- Péturaud, P.; Hampel, U.; Barbier, A.; Dreier, J.; Dubois, F.; Hervieu, E.; Martin, A.; Prasser, H.-M.,
General overview of the TOPFLOW-PTS experimental program
The 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-14, 25.-30.09.2011, Toronto, Canada
- Seidel, T.; Beyer, M.; Hampel, U.; Lucas, D.,
TOPFLOW-PTS air-water experiments on the stratification in the ECC nozzle and the ECC water mixing during PTS scenarios,
NURETH-14, 25.-29.09.2011, Toronto, Canada
- Apanasevich, P.; Lucas, D.; Hoehne, T.,
Numerical simulations of the TOPFLOW-PTS steam-water experiment,
The 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), 25.-30.09.2011, Toronto, Canada
- Lucas, D.; Bestion, D.; Coste, P.; Pouvreau, J.; Morel, Ch.; Martin, A.; Boucker, M.; Bodele, E.; Schmidtke, M.; Scheuerer, M.; Smith, B.; Dhotre, M. T.; Niceno, B.; Lakehal, D.; Galassi, M. C.; Mazzini, D.; D’Auria, F.; Bartosiewicz, Y.; Seynhaeve, J.-M.; Tiselj, I.; ŠTrubelj, L.; Ilvonen, M.; Kyrki-Rajamäki, R.; Tanskanen, V.; Laine, M.; Puustinen, J.,
Main results of the European project NURESIM on the CFD-modelling of two-phase Pressurized Thermal Shock (PTS),
Kerntechnik 74(2009), 238-242