[en] Pre-test calculations of the IAEA Standard Problem Exercise (IAEA-SPE-I) are presented. The experimental basis of the SPE is the PMK-NVH facility, the full-pressure 1:2070 scaled model of the Paks Nuclear Power Plant. The studied transient process was a 7.4% break in the cold leg of the reactor. The analysis was carried out by the KfK version of RELAP4/mod6 code. Major occurrances of the transient are given. (author)

[en] During the last few years the safety of the Hungarian Paks Nuclear Power Plant was reassessed in the framework of the AGNES project. Results of the program revealed that the safety of VVER-440/213 type reactors could be enhanced by modernizing a number of event oriented emergency operating procedures (EOPs) and the implementation of accident management (AM). Before the accomplishment of systematic AM all the possible thermal-hydraulic effects and consequences should be considered and experimentally verified. This paper summarizes the most important findings of the relevant tests performed in the PMK-2 facility, which is a full-pressure, integral-type model of the Paks NPP. The present analysis concentrates on a particular method, the bleed-and-feed, applied in the primary or secondary circuit and conclusions are drawn for the effectiveness of this AM measure. Modelling of the thermal-hydraulic processes is extended with computer simulations using the RELAP5/MOD3.2 system code. Finally, a short assessment is given for the code capabilities to represent some significant phenomena of the transients. (author)

[en] The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.)

[en] The RELAP5/MOD2 and MELCOR 1.8.1 codes have been used for simulate the PMK total loss of feedwater with secondary bleed and feed experiments done in a scale-model WWER-440 test facility. Nodalization studies and studies on several-core modelling options were also done. Good agreement was found between the calculations done by RELAP5/MOD2 and MELCOR 1.8.1 JY codes. (orig.) (5 refs., 31 figs.)

[en] Due to the specific features of the VVER-440/213-type reactors the transient behaviour of such a reactor system is different from the usual PWR system behaviour. To provide an experimental database for the transient behaviour of VVER systems the PMK integral-type facility, the scaled down model of the Paks NPP was designed and constructed in the early 1980's. Since the start-up of the facility 48 experiments have been performed. It was confirmed through the experiments that the facility is a suitable tool for the computer code validation experiments and to the identification of basic thermal-hydraulic phenomena occurring during plant accidents. High international interest was shown by the four Standard Problem Exercises of the IAEA and by the projects financed by the EU-PHARE. A wide range of small- and medium-size LOCA sequences have been studied to know the performance and effectiveness of ECC systems and to evaluate the thermal-hydraulic safety of the core. Extensive studies have been performed to investigate the one- and two-phase natural circulation, the effect of disturbances coming from the secondary circuit and to validate the effectiveness of accident management measures like bleed and feed. The VVER-specific case, the opening of the SG collector cover was also extensively investigated. Examples given in the report show a few results of experiments and the results of calculation analyses performed for validation purposes of codes like RELAP5, ATHLET and CATHARE. There are some other white spots in Cross Reference Matrices for VVER reactors and, therefore, further experiments are planned to perform tests primarily in further support of accident management measures at low power states of plants to facilitate the improved safety management of VVER-440-type reactors. (authors)

[en] The PMK-2 experimental facility at the KFKI-AEKI, Budapest, is a full pressure, scaled down model of the primary and partly the secondary circuit of the Paks Nuclear Power Plant. This NPP is equipped with four VVER-440/213-type reactors. Such plants are slightly different from PWRs of usual design and have a number of special features as 6-loop primary circuit, horizontal steam generators, loop seal in hot and cold legs, setpoint pressure of passive safety injection tanks (SIT) higher than secondary pressure, etc. The PMK-2 was primarily designed for investigating operational and off-normal transient processes, as well as small-break loss of coolant accidents of Paks NPP. The volume and power scaling ratios are 1:2070. Due to the importance of gravitational forces in both single- and two-phase flow the elevation ratio is 1:1 except for the lower plenum and pressuriser. The six loops of the plant are modelled by a single active loop. Transients can be started from nominal operating conditions. The pressuriser (PRZ) is connected to the lower part of the hot leg as in the reference system. The core model consists of 19 electrically heated rods. The main circulating pump of the PMK-2 serves to produce the nominal operating conditions and to simulate the flow coast-down following pump trip. The horizontal design of the VVER-440 steam generator is modelled by horizontal heat transfer tubes between hot and cold vertical collectors in the primary side. The emergency core cooling systems including the SITs. High and low pressure injection systems of the Paks NPP are also modelled. The first design of the PMK-NVH facility only modelled the primary circuit of plant. This version was used until 1990. The PMK-2 facility is an upgraded version (first of all by addition of a controlled secondary heat removal system) extending the capability of the test loop to modelling transient processes evoked by initiating events in the secondary circuit or including accident sequences in support of accident management (AM) procedures. During the 15 operational years - from May 1986 onwards with the first of four IAEA Standard Problem Exercise tests - 48 different experiments, including cold and hot leg break LOCA, primary-to-secondary leakage (PRISE), loss of flow, loss of feedwater, disturbances of natural circulation, etc. tests were performed on this integral type test facility. An overview on 11 experiments related to the operational behaviour of horizontal steam generators performed in framework of national research projects IAEA Technical Co-operation Project RER/9/004 (Standard Problem Exercises) and three EU PHARE projects (in co-operation with AEAT, FRAMATOM, SIEMENS, IPSN, GRS, FZR and VVER owner countries) is given in the first part of paper. In the second part results of two types of tests in shutdown condition with RELAP5 post-test analysis may be of interest to the computer simulation of the horizontal SG too - are summarised. (orig.)

[en] Analysis of postulated accidents in nuclear power plants are carried out by computer codes, which must be validated by experimental results. Experiments were carried out on the PMK-NVH facility, a scaled down model of the Paks Nuclear Power Plant equipped with WWER-440 reactors. This work presents the results of four experiments with cold and hot leg breaks modelling a 7.4% break in the plant starting from nominal operating conditions with and without hydroaccumulators in action and considering the availability of one high pressure system. A few results of the validation of RELAP4/MOD6 is also given in the report. (author). 7 refs and figs

[en] The PMK-2 experimental facility at the KFKI-AEKI, Budapest, is a full pressure, scaled down model of the primary and partly the secondary circuit of the Paks NPP, which is equipped with four VVER-440/213-type reactors. Since the start-up of the facility altogether 48 experiments have been performed for groups of transients as follows: one- and two-phase natural circulation, loss of coolant accidents, special plant transients and experiments in support of accident management procedures. The results have been used for the validation of thermal-hydraulic system codes for VVER applications. Following the experiments a detailed documentation and archiving activity - using an optimised data storage - was required to preserve the essential information and to assure these for a widely utilisation for the international nuclear community. In the publication list related to the facility and the experiments for the moment altogether 280 items - documents, articles in periodicals, papers in proceedings and research reports - in six languages were collected. The paper gives an overview on this activity including the participation in the EU CERTA-TN programme, where AEKI introduced representative databases of two PMK-2 tests in the STRESA Network.(author)

[en] Experiments have been performed with the PMK integral-type facility, a model of WWER-440 type PWRs, to investigate two-phase natural circulation behaviour. The phenomena to be expected in this reactor type are different from those in PWRs with vertical steam generators mainly due to the loop seal in the hot leg and the horizontal layout of the steam generator heat transfer tubes. The experiments showed that the system is repressurized when the water level drops to the hot leg elevation due to the effect of the loop seal. Opening of the loop seal can be smooth, but may lead to oscillations depending on the power and the mass inventory. Natural circulation recovers after the hot leg loop seal is opened, but then decreases with further mass inventory decrease. (orig.)

[en] At present, no systematic accident management has been implemented in Paks NPP of VVER-440/213-type nuclear power plant. However, results of the safety reassessment of the Paks NPP showed that one group of the measures to enhance the safety is the modernization of emergency operating procedures, including the implementation of accident management. Report is based on experimental results obtained on the PMK-2, an integral-type facility of the Paks NPP. The purpose is to study the application of secondary and primary side bleed and feed procedures to prevent the core damage for beyond DBA accident sequences. Report contains items as: assessment of the effectiveness of the operator interventions; performing accident management analyses for the PMK-2 facility; experimental results on the accident scenarios selected; computer code verification; assessment of the applied procedures