Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

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PKL is the only test facility in Europe that replicates the entire primary side and the most important parts of the secondary side of western-type Pressurized Water Reactors (PWR) in the scale of 1:1 in heights. It is also worldwide the only test facility with 4 identical reactor coolant loops arranged symmetrically around the Reactor Pressure Vessel (RPV) for simulation of nonsymmetrical boundary conditions between the reactor loops. Thermal-hydraulic phenomena observed in PWRs are simulated in the PKL test facility for over 40 years. The analyses carried out in these years encompass a large spectrum of accident scenario simulations and corresponding cool-down procedures. The overall goal of the PKL experiments is to show that under accident conditions - even for extreme and highly unlikely assumptions as additional loss of safety systems - the core cooling can be maintained or re-established by automatic or operator- performed procedures and that a severe accident e.g. a core melt-down can be avoided under all circumstances. Another goal of the tests performed in the PKL facility is the provision of data for validation of thermal-hydraulic system codes. This paper presents recent modifications of the PKL facility, applied in order to adapt the facility to the latest western-type designs currently built in the world. The paper discusses also important results obtained in the last years.