An integral reactor with forced coolant circulation at emergency power level operation and natural circulation for residual heat removal is used in the design of the VPBER-600 reactor plant for the power unit of a new generation NPP of 640 MW (el) power. ‘

Forced coolant circulation is provided with the help of six leak-tight circulation electric pumps, located on the bottom of the reactor vessel.

In the design of some equipment and systems the decisions have been mode which have been verified by long experience of operation of the existing nuclear power plants.

Calculation analysis of the wide range of accidents, performed on the base of both deterministic and probabilistic approaches demonstrated the high safety of the plant. Safety in the course of three days is provided by passive means without power supply from outside nor personnel intervention.

Form the point of view of a deterministic approach for severe core damage, multiple failures of safety systems elements and systems as a whole are necessary. The probability of severe damage to the core, evaluated deliberately conservatively is < 10’8 per reactor/year.

Nevertheless the search for design decisions for optimization of the reactor design and improvement of characteristics including safety provision for severe accident-accidents with postulated melting of the core continues.

As a result circulation pumps were moved from the bottom to the cylindrical part of the reactor vessel, reactor internal heat exchangers of the system for emergency heat removal were excluded, engineering decisions for the limitation of the consequences of severe accident were proposed and the possibility of corium confinement in the reactor vessel or guard vessel was shown.

Moving the circulation pumps to the cylindrical part of the vessel simplifies the operational servicing of the reactor, excludes the possibility of coolant leakage below the core and improves the conditions for corium confinement in the core and in the reactor vessel and for creation of an in-reactor corium catcher.

The schematic diagram of the system of severe accident localization, presented in Fig.3, includes:

— heat exchangers-condensers, of the system of purification and boric reactivity compensation, total power 10 MW are located in the guard vessel;

— two safety complexes (DN 100), each consisting of a membrane-rupture device and a safety valve in series;

— two temperature-actuated devices for the reactor pressure relief;

— bubbler of approximately 300 m3 volume;

— receivers of 600-700 m3 volume.

The bubbler and receivers have the same strength as the guard vessel. In a severe accident with core melting when the temperature in the reactor reaches 500-700°C, safe

SYSTEM OF SEVERE ACCIDENTS LOCALIZTION

Complex System of VPBER-600 Thermo-physics and Safety Investigations
Build-In steam-gas pressorizer
Investigation* of dynamics at normal and emergency conditions
Investigations of Static dtrtnbuUont of temperature* and gas
О as transfer processes investigations
Passive system of heat removal
Investigations of the efficiency of dry fleam generator In steam and steam-gas mixture
Investigation of tesmal efficiency of heat-exchanger-condenser in ateam-gaj medium
Power determinations, trying-out of start up conditions in water
Loss of tightness accidents
Investigation of termo — hydndic processes in the reactor system-guard vessel	Determination of the efficiency of sbtems/ influencing upon the accident run (system of passive heat removal, hydraulic accumulator	InveshgiiKms of icihi hydraulics of primary «.псин at tuiural circulation and loss of circulation
Hydrodynamics of the core Inlet chamber
Hydrodynamics investigation and development of live structure for rates eguabzation at core inkt
Determination of temperatures variations at partially disconnected steam generators
Complex Investigation of the KP at normal and emergency conditions
Aerodynamic model
Severe accidents with the core failure
Cooling of the reactor Investigation In accordance Cooling of the core vessel and guard vessel with complex program on severe accident* in steam gas medium llc. il removal
Fig.4

temperature devices of the reactor are opened and steam-gas mixture is discharged from the reactor to the guard vessel, untilpressure is equalized in the guard vessel and the reactor. When pressure in the guard vessel is 5.0 MPa a membrane-rupture device on the guard vessel is broken, a safety valve is opened and steam-gas mixture is discharged from the guard vessel to the bubbler. Gases liberated in core melting are pressed out to receivers, this passively solves the problem of provision for hydrogen safety.

Heat removal from the reactor vessel when corium is confined in the reactor or from the guard vessel false bottom when corium is confined in the guard vessel if it leaves the reactor vessel is performed with the help of heat exchangers-condensers in the guard vessel of the purification system. —

In normal operation heat exchangers in the guard vessel of the purification system are disconnected from heat exchangers unit by the pipeline for water supply and are connected remotely by the operator in the event of severe accident.