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The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection.
The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation.
Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society.
The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation.
Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies.
Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th).
The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation.
The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology.
Investigation of Corrosion Resistance of Alloys with Potential Application in Supercritical Water-cooled Nuclear Reactors