Investigating the thermal hydraulics of VVER-Type reactors in emergency and transient modes of their operation at the PSB VVER test facility

Abstract

In 1998, the PSB-VVER integral test facility simulating the primary coolant circuit of the VVER-1000 reactor plant with a power capacity and volumetric scale of 1:300 was constructed and put in operation. At present, PSB-VVER is the largest test facility able to operate at the nominal coolant parameters, which contains, to the fullest extent, all main elements of the primary coolant circuit, including the safety systems. The PSB-VVER test facility is intended for obtaining experimental data necessary for verifying the thermal hydraulic codes aimed at analyzing the thermal hydraulic processes in VVER-based reactor plants. In addition, the PSB-VVER test facility allows the researchers to analyze methods and algorithms to control the reactor plant under the conditions of anticipated operational occurrences and in emergency modes of its operation. It also offers the possibility to experimentally confirm the operability of power unit elements and systems in various modes of their operation and to experimentally verify the engineering solutions proposed for the new designs of reactor plants. The PSB VVER integral test facility is briefly described, and the experiments carried out on it are briefly reviewed. The use of the experimental data obtained at the facility for verifying the RELAP5 code is illustrated taking as an example the "4.1% leak from the cold leg" experiment carried out within the framework of the OECD-sponsored project. The use of the experimental facility for checking the engineering solutions proposed in the new designs of reactor plants and the use of experimental data obtained at the PSB-VVER test facility for analyzing the reactor plant control methods and algorithms under the conditions of anticipated operational occurrences and during accidents are considered on the examples of substantiating the new passive safety systems applied in the AES-2006 and VVER-TOI standard NPP designs (the latter abbreviation stands for a standard optimized design of an NPP based on the technology of VVER reactors with enhanced support by information technologies), as well as on the examples of studies carried out within the framework of the TACIS project devoted to checking the procedures for managing accidents at NPPs equipped with VVER-1000 reactors. The facility operation results are briefly summarized and the lines of its further use are outlined.