TECHNICAL PROBLEMS OF CREATION OF POWER REACTORS COOLED BY WATER OF SUPERCRITICAL PRESSURE

Abstract

The supercritical Water-cooled Reactor (SCWR) was adopted as one of the promising IV-generation reactors within the framework of the international forum “Generation-IV” (MFP). More than 45 organizations in 16 countries with advanced nuclear power are developing SCWR concept proposals for this program. The SCWR concept is based on the implementation of a direct-flow single-circuit scheme of a nuclear power plant, cooled by SCP water. The introduction of this type of nuclear power plant will increase the efficiency up to 45 %, increase the fuel reproduction coefficient, reduce metal consumption and construction volumes, and improve economic and environmental performance. Countries participating in the SCWR MFP consider the development of a reactor with a thermal neutron spectrum and uranium fuel as a priority task, but in the subsequent stages, with increasing problems with the storage of spent nuclear fuel (SNF) and small actinides (SA), it is possible to switch to a reactor with a fast neutron spectrum, MOX fuel and a closed fuel cycle (CFC). Within the framework of the MFP, various versions of SCWR are being developed differing in the parameters of the coolant and its circulation schemes in the core. Groups have been created to study the issues of physics, thermohydraulics, heat transfer, materials, personnel training. Water-cooled reactors research carried out during ~15 years in A.I. Leypunsky Institute for Physics and Power Engineering (IPPE), OKB “Gidropress”, NRC “Kurchatov Institute” with supercritical thermal and fast neutron spectra, it seems more promising to develop a reactor with fast spectrum of neutrons. For ~10 years, IPPE and OKB “Gidropress” have been working together on the VVER-SKD concept project - a single-circuit RC with a coolant SCP with a fast-resonance neutron spectrum with a capacity of Ne = 1700 MW. This rector is recognized as a prospect for the development of VVER technology with the possibility of using uranium fuel and switching in the future to MOX-based fuel (U-Pu-Th) and to SNF. When developing VVER-SKD, it is necessary to solve a complex of scientific and technical problems: development and verification of calculation codes of improved estimation for neutron physics, hydrodynamics and water heat transfer of SCP in fuel assemblies (FA) of the core and throughout the reactor; development of fuel elements and FA structures, justification of their operability; analysis of reactor stability under transient and emergency conditions; selection of heat-resistant structural materials for fuel rods and reactor vessel with high corrosion resistance; justification and development of optimal water-chemical regime, etc. Some of these problems are investigated in bench and loop tests, but to solve most of them and justify the technology for subsequent licensing, it is necessary to create an experimental test reactor. In relation to the VVER-SKD reactor Ne = 1700 MW, the paper presents the results of calculations of fuel cycles with MOX and nitride fuels, justifies the use of NPS in the ZTC, discusses the problems of heat exchange and thermal hydraulics, and suggests a test reactor option.