No abstract

In the PSB-VVER Integral Test Facility (ITF), extensive experimental investigations on the thermal hydraulic behavior of a VVER-1000 type reactor under LOCA accidents have been carried out during 2001. The general aim of the experiments is to contribute to a better understanding of accident sequences and to provide a detailed data base for the validation of Russian and Western computer codes. This paper presents the results of an 11% cold leg break experiment (CL-11-05) assuming loss of off-site power. The presented experiment belongs to a set of tests with a spectrum of break area in the PSB cold leg carried out in the summer 2001. The results of the post-test analyses using TRAP code are presented. A short description of the TRAP code mathematical model is also given.

New design of nuclear power plant with pressurized water reactor entitled “NPP-2006” is under progress in Russia. It is evolutionary type of design based on existing NPP with VVER-1000 reactors. Intensive thermal-hydraulic experimental activities support design development. They are aimed at increasing of NPP safety and efficiency. Integral test facility PSB-VVER, modeling main components of primary system of VVERs, is the main tool for experimental studies. Volumetric-power scale of the test facility is 1:300, elevations are 1:1. A core model is an assembly of 168 fuel rod imitators. About 20 experiments at the PSB-VVER have been performed to generate experimental data base for validation of best estimate thermal-hydraulic system codes for application to analyses of VVERs transients and accidents. An important direction of PSB-VVER experimental activities is connected with the development and verification of accident management procedures. 15 experiments have been conducted to support development of such procedures. For new plant design (hydroaccummulators, operating under low pressure) is under experimental studies at the PSB-VVER. The experimental program also includes the approach, based on investigations of local thermal-hydraulic parameters in the fuel rod imitators assembly at the large-scale test facility. This approach is supposed to decrease a level of design conservatism and to enhance NPP efficiency.

Results of experiments conducted on a large-scale integral PSB-VVÉR bench and directed toward investigation of the performance of new passive safety systems for nuclear plants with a water-cooled power reactor, an additional system for passive flooding of the active zone, and a system for the passive removal of heat from steam generators are presented.Keywords: nuclear power plant, water-cooled power reactor, passive safety system.A new generation of "AÉS-2006" nuclear plants with a water-cooled reactor is currently under development in Russia. This is an evolutionary design in which experience gained with the design, construction, and operating experience of active nuclear power plants (NPP) equipped with the VVÉR-1000 reactor unit is reflected. To prevent the occurrence of a broad spectrum of projected emergencies with loss of heat-transfer medium and failure of active safety systems in a critical stage, the new design calls for use of the following modern passive safety systems: a supplementary system for passive flooding of the active zone (SSPFAZ) and a system for passive removal of heat from the steam generators (SPRH SG).In the absence of full-scale field investigations, the behavior of reactor units (RU) together with modern passive systems is currently modeled by mathematical means using systems thermohydraulic codes, and, correspondingly, the reliability of preventing a serious emergency will be directly dependent on the reliability of the methods employed for design analysis.The difference in the results of analyses performed for different codes is associated with the fact that the SSPFAZ and SPRH SG are new systems operating in a range of thermophysical parameters for which the codes were not initially intended.To substantiate the effectiveness of the new passive safety systems, and acquire experimental data for verification of the systems thermohydraulic codes on a large-scale integral PSB-VVÉR bench, we conducted experiments simulating an emergency with the guillotine severing of a "hot" pipeline and operation of the passive SSPFAZ and SPRH SG safety systems.