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. 1. Introduction Complex thermal-hydraulic system codes are used for the analysis of accident sequences in pressurized water reactors. The necessity to verify the knowledge gained using such codes by experiments in suitable test facilities resulted in the construction of the large-scale test facility PKL (from the German abbreviation for Prim？rkreislauf) modeling a 1300？MW class PWR. The PKL test facility has been in operation since 1977; however, in the meantime, the objectives of the experiments performed at the PKL test facility have changed considerably with the result that the test rig has been refitted many times to suit the additional and ongoing tasks and also to match latest developments, for example, in the fields of measuring instrumentation and data processing. Since the commencement of experiments at the PKL test facility, the various phases of the experiments have always reflected and given priority to current safety issues. The primary objective of all PKL experiments has been and remains the experimental investigation of thermal-hydraulic processes in PWRs with respect to the response of the overall system. To some extent the investigations also include the behavior of individual components and subsystems during the simulation of operational transients and accidents. The tests performed to date (in total more than 150 integral experiments) have altogether
References
[1]
R. Mandl, B. Brand, and H. Watzinger, “PKL reflood tests including end-of-blowdown,” in Proceedings of the 13th Water Reactor Safety Research Information Meeting, Gaithersburg, Md, USA, October 1985.
[2]
R. M. Mandl and P. A. Weiss, “PKL tests on energy transfer mechanisms during small-break LOCAs,” Nuclear Safety, vol. 23, no. 2, pp. 146–158, 1982.
[3]
K. J. Umminger, W. Kastner, R. Mandl, and P. Weber, “Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: conclusions from an experimental program in a 4-loop test facility (PKL),” in Proceedings of the 2nd ASME-JSME Nuclear Engineering Joint Conference (ICONE '93), vol. 1, pp. 409–416, San Francisco, Calif, USA, March 1993.
[4]
R. Mandl and B. Brand, “PKL III small breaks and transients experimental programme,” in Proceedings of the 14th Water Reactor Safety Research Information Meeting, Gaithersburg, Md, USA, October 1986.
[5]
J. Liebert, B. Brand, W. Schwarz, G. Sgarz, and K. Umminger, “Results of UPTF and PKL research projects for PWR plant operation,” VGB PowerTech, vol. 79, no. 8, pp. 20–27, 1999.
[6]
Y. Anoda, J. Katayama, Y. Kukita, and R. Mandl, “Secondary bleed and passive feed during PWR station blackout (TMLB') transient: experimental simulation at full pressure and temperature,” in Proceedings of the 113th ASME Annual Winter Meeting, pp. 89–96, Anaheim, Calif, USA, November 1993.
[7]
B. Brand, R. Mandl, and H. Watzinger, “Investigation of PWR transients in the PKL test facility,” in Proceedings of the 3rd International Topical Meeting on Nuclear Power Plant Thermohydraulics and Operations (NUPTHO '88), Seoul, Korea, November 1988.
[8]
K. J. Umminger, W. Kastner, R. Mandl, and P. Weber, “Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: conclusions from an experimental program in a 4-loop test facility (PKL),” in Proceedings of the 2nd ASME/JSME International Conference on Nuclear Engineering (ICONE '93), pp. 409–416, San Francisco, Calif, USA, March 1993.
[9]
K. Umminger, S. P. Schollenberger, S. Cornille, C. Agnoux, D. Quintin, and P. Freydier, “PKL tests on heterogeneous inherent boron dilution following Sb-Loca—applicability to reactor scale,” in Proceedings of the 18th International Conference on Nuclear Engineering, Proceedings (ICONE '10), vol. 4, pp. 433–439, Xi’an, China, May 2010.
[10]
K. Umminger, B. Schoen, and T. Mull, “PKL experiments on loss of residual heat removal under shutdown conditions in PWRs,” in Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP '06), pp. 1776–1784, Reno, Nev, USA, June 2006.
[11]
K. Umminger, L. Dennhardt, and S. Kliem, “Experiments on main steam lineBreak in the test facilities PKL and ROCOM,” in Proceedings of the 14th International Topical Meeting on Nuclear Thermalhydraulics (NURETH '11), Toronto, Canada, September 2011.
[12]
A. Del Nevo, E. Coscarelli, A. Kovtonyuk, and F. D'Auria, Analytical Exercise on OECD/NEA/CSNI PKL-2 Project Test G3.1: Main Steam Line Break Transient in PKL-III Facility, Pisa, Italy, 2010.
[13]
F. Reventós, J. Freixa, L. Batet et al., “An analytical comparative exercise on the OECD-SETH PKL E2.2 experiment,” Nuclear Engineering and Design, vol. 238, no. 4, pp. 1146–1154, 2008.
[14]
A. Bucalossi, Validation of Thermal-Hydraulic Codes for Boron Dilution Transients in the Context of the OECD/SETH Project, EUROSAFE, Brussels, Belgium, 2005.
