PMK2-VVER440 Reports:†††††††††† Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies.
MTA KFKI Atomic Energy Research Institute, Budapest, Hungary and Akademiai Kiado Budapest Hungary, through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France.
PDF format; many computers (M00012MNYCP00).
The PMK-2 facility is located at the KFKI Atomic Energy Research Institute (AEKI), Budapest, Hungary. It is a full-pressure thermohydraulic model of the primary and partly the secondary circuit of the Paks nuclear power plant of VVER-440/213 type. At the start-up time in 1985 PMK-2 was the first and the only integral-type facility for VVERs. It was designed and constructed to aid in the understanding of system behaviour and to provide databases for computer code validation. The PMK-2 was followed by the PACTEL facility for VVER-440 in Finland (1990) and the ISB and PSB facilities for VVER-1000 in Russia (1992 and 1998, respectively).
Since the start-up of the PMK-2 facility in 1985, altogether 55 experiments have been performed primarily with the participation of several international experts from European and other countries to study one- and two-phase natural circulation, loss-of-coolant accidents (LOCA), special plant transients and accident management (AM) procedures. The results have been used for the validation of thermohydraulic system codes like ATHLET, CATHARE and RELAP5 for VVER applications.
A large number of integral-type tests have been performed in the past 30 years all over the world in different test facilities. In the last decade the OECD CSNI recognized that there is a tremendous danger of losing these test results due to the closing of many of these facilities, the retirement of experts and changes in data storage. A huge effort was initiated in the 90s to collect the most important data included in the CSNI Integral Test Facility Validation Matrix at the NEA Data Bankówith limited success. In order to avoid this problem with respect to the PMK-2 data, the present books summarize the results and major findings of experiments.
Volume I contains the description of design features of the PMK-2 facility, modelling of the VVER-440/213 specific design solutions, controls and actions for safety systems. The 55 experiments performed cover an almost complete spectrum of design basis accidents included in the Safety Analysis Report of the Paks plant. The OECD-VVER cross reference matrices are developed for the PMK-2 tests for large breaks, small and intermediate leaks and transients, providing internationally accepted methodology for the identification of major phenomena addressed by the tests.
Volume II provides the identification and discussion of major findings of experiments, in terms of the OECD-VVER code validation matrices. It also presents the validation results of different versions of the ATHLET, CATHARE and RELAP5 codes applied to safety assessments in Hungary. The validation covers both the conventional qualitative method and a quantitative method based on fast Fourier transform.
The PMK-2 database has significant value for the safety evaluation of the VVER-440/213 type reactors in operation. The database provides a valuable source of scientific information to the nuclear community in the world. The 20 year operational period established an international scientific school for experts of 29 countries participating in the PMK-2 projects.
Adobe Acrobat Reader can be used to access the files in Portable Data Format (.pdf).
L. Szabados, et al. Final Report on the PMK-2 Projects-Volume I, Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies, (Volume I), ISBN 978-963-05-8461-6, 236 pages with a CD supplement, which contains the experimental results in digital form, containing 2640 files.
L. Szabados, et al. Final Report on the PMK-2 Projects-Volume II, Major Findings of PMK-2 Test Results and Validation of Thermohydraulic System Codes for VVER Safety Studies, ISBN 978-963-05-8810-2 (2009-11-15).
The package is transmitted on a CD with the books in PDF format, related data and documentation.
KEYWORDS: REACTOR SAFETY, LOCA, THERMAL HYDRAULICS