The experiment to the IAEA standard problem exercise No. 4 was carried out in April 1993 on the integral test facility PMK-2 in Budapest. It was a 3.2 mm break on the downcomer head. The high pressure injection cooling was assumed to be not available. As an accident management measure bleed and feed on the secondary side of the steam generator was applied. Research Center Rossendorf contributed to the experiment of SPE-4 by supplying needle shaped conductivity probes for the measurement of local void fractions in the primary circuit of the PMK-II test facility. In the course of the standard problem exercise No. 4 RCR contributed with posttest calculations using the thermalhydraulic code ATHLET. The report comprises a description of the init...
Within the framework of the arrangement on the safety research participation and the technical excha...
Due to the character of the original source materials and the nature of batch digitization, quality ...
An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LST...
The experiment to the IAEA standard problem exercise No. 4 was carried out in April 1993 on the inte...
The experiment to the IAEA standard problem exercise No. 4 was carried out in April 1993 on the inte...
In the framework of the computer code assessment programme for the VVER-440 type Paks Nuclear Power ...
The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the P...
ABSTRACT The PWR PACTEL test facility has recently been designed to support the safety studies of EP...
Temperature gradient on the thick Reactor Pressure Vessel (RPV), caused by sudden overcooling events...
In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to su...
International audienceIn the course of a LOCA accident, the fuel rod can be deformed and thus, it is...
The OECD/NEA PSB-VVER project provided unique and useful experimental data from the large-scale PSB-...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
The document deals with the description of results obtained by the Relap5 code in the simulation of ...
Within the framework of the arrangement on the safety research participation and the technical excha...
Due to the character of the original source materials and the nature of batch digitization, quality ...
An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LST...
The experiment to the IAEA standard problem exercise No. 4 was carried out in April 1993 on the inte...
The experiment to the IAEA standard problem exercise No. 4 was carried out in April 1993 on the inte...
In the framework of the computer code assessment programme for the VVER-440 type Paks Nuclear Power ...
The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the P...
ABSTRACT The PWR PACTEL test facility has recently been designed to support the safety studies of EP...
Temperature gradient on the thick Reactor Pressure Vessel (RPV), caused by sudden overcooling events...
In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to su...
International audienceIn the course of a LOCA accident, the fuel rod can be deformed and thus, it is...
The OECD/NEA PSB-VVER project provided unique and useful experimental data from the large-scale PSB-...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
The document deals with the description of results obtained by the Relap5 code in the simulation of ...
Within the framework of the arrangement on the safety research participation and the technical excha...
Due to the character of the original source materials and the nature of batch digitization, quality ...
An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LST...