A nuclear power plant with a sodium-cooled fast reactor is considered safer and less expensive if its primary (activated) sodium circuit has an integral arrangement, with the heat from the sodium transferred to a steam-water circuit by an intermediate steam generator circuit with single-wall tubes. However, these advantages can at the same time be regarded as disadvantages.

For example, the Beloyarsk BN-600 is such an integral type fast reactor1 (see figure). The reactor’s primary circuit equipment (which includes three main coolant pumps (MCPs) and six intermediate heat exchangers (IHXs) with a total mass of 650 t), is located in the reactor vessel around the core and rests on complex massive in-vessel structures. As a result, the reactor size and metal content are very large. In addition, the reactor was constructed at the power plant site, where some technical requirements could not be met. The primary equipment is activated during reactor operation which impairs its maintainability. Furthermore, the dismantling of this cumbersome vessel will be a complex technical problem.

The BN-600’s intermediate three-loop sodium circuit is equipped with section-type steam generators. The steam generator in each loop consists of eight sections, with each section including evaporator, superheater and intermediate superheater. Numerous valves are installed in the sodium and water circuits to control sodium and water flow through the sections. Heat transfer in the steam generator is carried out in single-wall tubes, and thus a possibility of dangerous sodium-water interaction can not be eliminated.

The authors set themselves a task to develop a more reliable and safer sodium cooled fast reactor power plant. The reactor will have a smaller metal content and lower capital construction cost, a simpler heat flow scheme and, therefore, better economic parameters.


The nuclear power plant unit proposed includes a compact fast reactor with loop-type sodium circuit and an air auxiliary system for air cooling of the reactor vessel, or RVACS.2 The cooling circuit equipment is located outside the reactor and, therefore, the reactor diameter and metal content will be less, and the structure simpler than an integral-type reactor. A loop-type reactor can be manufactured at the fabrication plant and transported to the nuclear power plant site. Thus, the manufacturing quality will be higher and the ability to transport the unit assembled will simplify its dismantling and removal after the end of its life. The reactor diameter and weight can be chosen in accordance with existing hoisting and transport means if there are physical restrictions. The reactor diameter determines the core diameter and reactor design power. Taking into account the transport means available, one can consider the reactor thermal power to be in the range from 800 to 1000 MWe and close to the passive PRISM reactor module.2 The reactor vessel (shown on the following page) is surrounded by a protective “guard” vessel, the gap between them being filled with argon. A skirt inside the reactor vessel is connected to the cover dividing the internal space into pressure and discharge plenums; the vessel contains a “compensation” volume in the circuit above the level of the sodium, and is equipped with self-actuating valves below the compensation volume. Inlet and outlet branch pipes are provided to connect the reactor with the sodium circuit. Outlet branch pipes are lowered into the discharge plenum below the compensation volume. The latter is filled with argon whose pressure exceeds that of the sodium column above its level in the compensation volume.

When choosing the vessel height and diameter, an additional requirement must be considered, namely, the provision of emergency core cooling by the natural sodium circulation between the core and the inner surface of the cooled vessel. Heat from the reactor vessel is transferred through the gap filled with argon to the guard vessel. The RVACS system will provide the guard vessel cooling with the atmospheric air naturally circulating in the reactor vault. Similar to the PRISM design, the RVACS is continuously in operation and does not require operator intervention.

For heat transfer from the sodium circuit to the steam-water circuit, shell-type steam generators are used similar to that patented by the author, N A Ermolov3. In these, sodium and water are enclosed in heat transferring and steam-generating panels, respectively, and thus are separated by two walls and an intermediate cavity. Heat transfer between panels is carried out by helium convection and heat conduction and panel heat radiation. The sodium and steam-water panels contain long tubes (90 m) in a spiral form.

Preliminary calculations have shown that a heat transfer surface area of the heat exchanger, where helium convection and heat conduction as well as radiation through helium are used, is 5-6 times more than that of liquid sodium heat exchanger. This value can be reduced approximately 1.5-2 times by modern methods for intensification of convective heat transfer. Inlet and outlet temperatures for the core and steam generator equal to those in BN-600 were adopted as initial data for comparative calculations. A more exact ratio of surface areas and working helium pressure can be obtained by an optimisation thermal-hydraulic calculation.

The steam generators are located above the reactor: to be able to take advantage of the natural circulation of sodium; to make the unit more compact; and to improve the sodium circuit structure. With this lay-out, locations with increased thermal stresses and equipment component deformations can be fully eliminated in the sodium circuit. An essential saving of pumping energy can be obtained through natural circulation of sodium. The height of the sodium column above its level in the compensation volume will be about 30-35 metre, and helium pressure in this volume – about 0.35 MPa. At this height, the distance between the middle line of the core and the steam generator can be about 35 m, and a hydrostatic pressure of the cold downward circuit part will be about 0.025 MPa.

A forced sodium circulation is provided by MHD-pumps, which increases leak-tightness, reliability and safety of the circuit. One freezing valve and one quick-response stop device are installed in each loop of the main sodium circuit. A quick- response stop valve is also installed in each loop of the main steam-water circuit.

The unit is completely maintainable. The sodium circuit components are not activated due to their location outside the reactor. A loop with failed equipment can be isolated, repaired and put into service again without interrupting the reactor operation. A compact equipment lay-out allows almost the entire unit to be located on a seismically isolated platform inside the containment. The reactor vessel has no penetrations; therefore, sodium remains in the vessel even in case of a full ruputure of the main piping.

The skirt dividing the internal space into pressure and discharge plenums, self-actuating valves, sufficient height and vessel surface area are all important features of the structure that ensure a reliable core cooling after emergency reactor shutdown by natural circulation of sodium and circulation of air by the continuously operating RVACS system.


One possible reactor structure is shown in the following two diagrams, with the number of loops chosen at random.

The reactor as shown includes the vessel (3), the cover (14) and, enclosed in the protection shell (4) with a gap (5) filled with argon, the core (2), the skirt (7) and the radial radiation shielding (6). The cover is equipped with a rotating plug (20) with a built-in rotating column (19). The skirt is connected to the reactor cover, dividing the sodium-filled space into pressure cavity (36) and discharge region (8) which contains the compensation volume (15) filled with argon, in the sodium circuit. The reactor is connected to the sodium circuit by inlet and outlet branch pipes (10), with the outlet branch pipes placed below the compensation volume. The skirt is equipped with normally closed self-actuating valves (9) below the compensation volume. The reactor is arranged in the vault (1), which forms part of the reinforced concrete base (11). The basic sodium and steam-water circuit equipment is located in four boxes (38) above the reactor and in channels connected thereto. The rooms shown (23) are designed for housing of auxiliary system and repair zone equipment.

Each box (38) contains main piping (28 and 31) of the sodium circuit with the fast-response stop devices (32) and freezing valves (29), MHD-pump and shell type steam generator (25). Each steam generator is made up of the heat exchanging panels (27) and steam generating panels (26) of the sodium and steam-water circuits, respectively, and the intermediate helium cavity between the panels. The intermediate cavity is connected by pipe (33), equipped with a large diameter safety rupture device (34) and stop valve (35) with a cooled large volume reservoir with radiation protection. The shell is designed to promptly receive sodium-water reaction products from the steam generator in case of an incident; a remote-controlled system is included for their removal.

Each steam generator is equipped with a relief valve (24) for emergency helium pressure drop in the intermediate cavity in case of loss of leak tightness in a sodium circuit panel. Main water-steam pipes (40) are equipped with fast-response devices (39) located in the rooms (23). The main sodium and steam-water pipes are connected with the corresponding panels through pressure and discharge collectors located outside the steam generator (not shown in the figures). The RVACS (22) is constantly cooling the vessel (3), blowing air circulating in the vault, down the guard shell (4).

The unit includes a box (18) with equipment for fuel reloading. The reactor drainage cavity is connected to the equipment box by an elevator (16). The RVACS system and reactor boxes with the steam generator panels and fuel reloading equipment, which are covered by the containment (17) and located below the ground elevation, are installed on the platform (13). The platform is seismically isolated from the reinforced concrete basement by mechanical isolators (12).


The unit operates as follows: During reactor operation, the seals of the rotating plug (20) and rotating column (19) are frozen and an excessive argon pressure is maintained in the compensation volume (15) to compensate the static pressure of sodium vapour present above its level in the compensation volume. MHD-pumps (30) installed in the main piping (31) supply cooled sodium to the reactor annular pressure cavity (36). The sodium goes down through the annular pressure cavity, cooling the vessel (3) and radial radiation shielding (6), passes the skirt and goes through the core (2) and after that to the discharge cavity (8). Sodium passes through outlet branch pipes (10) to the main piping (28) and goes to the upper part of the steam generator (25). Passing through the sodium panels (27), the coolant transfers heat to the steam generating panels (26) by helium convection and heat conduction in the cavity between the panels and by heat radiation and is then supplied to the pump (30) inlets.

Emergency core cooling system

Immediately after a reactor emergency shutdown due to the loss of sodium flow in the main circuit the self-actuating valves (9) open. If sodium is retained in the circuit (ie, no leaking), two natural circulation paths are created. One part of the sodium will naturally circulate through valves in the skirt between the core and the inner reactor surface. The other part of the sodium will naturally circulate in the standard forced circulation circuit. Sodium is cooled in the steam generator, goes down to the annular pressure cavity, then to the core and again to the steam generator. Naturally circulating sodium will transfer heat to the steam-water circuit and to the air of the continuously operating passive RVACS system (22) and thus cool the core. In case of rupture of one or more of the main pipes, sodium will leak from the circuit into channels (21) and boxes (38), remaining in the reactor. In this case emergency core cooling will be carried out by sodium naturally circulating only in the inner circuit. Sodium will remove heat from the core and transfer it to the vessel wall. Heat from the wall will be transferred through the cavity (5) filled with argon to the shell (4) by heat radiation, as well as by argon convection and heat conduction. Heat from the outer shell surface will be removed by air naturally circulating in the vault.


To remove one loop for repair while the other loops remain operating, it is necessary, changing the sodium flow by MHD pumps, to stop sodium circulation in the loop, close the fast-response stop devices (32) in the main piping, and after that a small sodium flow can be established through clearances in the stop valves. Regulating the pump capacity, one can establish zero flow (discontinue the circulation) in the loop again, feed cooling medium to valves (29), and freeze the sodium in them.


For refuelling, the reactor is shutdown, the circuit loops are isolated from the reactor, argon pressure in the compensation volume is reduced, and the seals of the rotating plug (20) and rotating column (19) are heated up. During fuel reloading the core is cooled by sodium naturally circulating inside the reactor and by air from the passive RVACS system (22). The containment (17) and reinforced concrete walls of the boxes (38) located below the ground elevation, protect the power unit from diversion, radioactive releases and sodium fire. The protection from earthquakes is performed by isolators (12), installed between the platform (13) and reinforced concrete base (11).


Many safety problems of the proposed design have been considered and can be resolved, assuring safety in the core, sub-assemblies, fuel pins, control rods, control systems, fuel loading mechanisms, radiation shielding, cooling of shutdown reactor (by naturally circulating sodium and air), seismic isolation of the unit, auxiliary systems, etc.

A distinguishing feature is that the reactor operates under relatively low pressure of about 0.9-1.1 MPa. In particular, this low pressure provides for a substantial simplification of the reactor structure.

Another important feature is the use of steam generators in which the panels containing radioactive sodium are separated from the steam generating panels by an intermediate cavity filled with circulating helium under pressure. Both the sodium and steam-water panels are manufactured from long spiral tubes which can compensate for any deformations, increasing steam generator reliability noticeably.

A low concentration of chemically aggressive impurities in helium can be maintained in the intermediate cavities of all the steam generators and thus the corrosion of panel external surfaces is eliminated. Because helium is an inert gas, it is possible to use tubes made from inexpensive carbon steel in the steam generating panels. In addition, the inner tube surfaces will be less subjected to corrosion, because the aggressive impurities in the sodium can be maintained at a low level due to the leak-tightness of the sodium circuit and the helium cavity between the circuits. The steam generators also have other positive reliability features.

Nevertheless, it would be appropriate to consider effects of anticipated failures in the steam generator on unit safety.


In case of one or more tube ruptures in the sodium panels, helium will enter through the openings to the sodium circuit. If no preventive actions are taken, helium will be mixed with sodium in the main circuit and enter the compensation volume. In this case, various changes will be observed such as a decrease in the helium pressure in the steam generator and in sodium flow in the loop, an improvement of heat transfer in the core, a pressure increase in the compensation volume and variations in reactivity.

A drop in sodium flow in the loop will be caused by a decrease in filling of MHD-pump cross-section by sodium dilution by helium bubbles.

A certain cooling of the core immediately after the accident begins occurs which can be explained by the presence of helium bubbles in sodium. It was shown in reference 4 that the buoyancy force exchanges water flow turbulence and increases the heat transfer coefficient. The hydrodynamics of sodium is essentially similar to that of water; therefore, the increase in turbulence and heat transfer will be observed with the appearance of gas bubbles in sodium.

The variation of reactivity will be also caused by the “dilution” due to helium bubbles. This reactivity effect in BN-type reactors is usually associated with gas pressure variations in the reactor gas plenum; it is termed the barometric effect. In small reactors this effect is positive. In power reactors this effect is negligible; for example, in the BN-350 reactor it has not been observed. In case of the passing of large gas bubbles through the core, a sodium void reactivity effect (SVRE) can appear, its value also being dependent on the core size. The reactivity effect caused by sodium removal from the central sub-assembly of a BN-600 reactor is – 2 x 10-5, while that for a larger BN-1600 reactor it is + 8 x 10-5. The full core voiding results in the reactivity effect of – 0.01 and + 0.015, respectively1. As the proposed reactor core will have smaller dimensions compared to that of the BN-600, it can be expected that the SVRE for this reactor will also be negative. Absolute values of reactivity effects for different cases of helium entering into the sodium circuit can be determined in the course of core calculations.

Immediately after the appearance of a helium leak into the sodium circuit, the relief valve (24) will be automatically opened and the excessive pressure in the intermediate cavity will decrease. Then, the reactor power must be reduced, and the loop with the failed steam generator, isolated and removed for repair.

In case of the steam-water circuit tube rupture in a steam generator, pressure in intermediate gas cavity will increase. Immediately thereafter, the safety rupture device (34) opens, the fast-response valves (39) in the main steam-water circuit piping (40) close, and steam will enter through pipes (33) into the protected cooled reservoir (37) and condensed in it. The valve (35) is constantly open and serves only for separating the reservoir when replacing a ruptured device. The failed loop can also be separated for repair.

The accident with simultaneous rupture of steam-water and sodium tubes seems to be of low probability, as the power unit will be equipped with the means for early accident detection. Nevertheless a possibility for prevention of such an incident is provided in the design. The development of an accident will depend on the pressure difference between the steam-water circuit and helium in the intermediate cavity of the steam generator. One can expect that this accident will not proceed intensively due to the initial mixing of water steam and helium, the long tubes of the sodium panels and the long length of main piping. Therefore, due to fast response of the stop and rupture devices, it will be possible to localise the accident in a steam generator and release the reaction products into large volume protected cooling cavity (37). The dynamics of such an accident can be experimentally studied in advance.


The sodium circuit loop lay-out described above makes use of sodium cooled, natural circulation shell-type steam generators in which heat exchanging panels provide the transfer of the heat from the radioactive sodium to steam-generating panels by the convection and heat conduction of highly turbulent helium flow under pressure and by radiation. This design achieves high levels of safety, reliability, and economical efficiency for a power station with a fast reactor.

Helium as the intermediate heat transfer medium can be replaced by other less expensive coolants, such as, for example, another inert gas or liquid metal. Helium was chosen mainly due to its inertness and satisfactory thermal-physical properties. Furthermore, helium provides additional possibilities for early leak detection in steam generators.

The heat flow scheme of this proposed design of a sodium cooled reactor can be used for a reactor cooled by molten-salt. The working temperature of a molten-salt coolant is higher than that of sodium, and heat transfer by radiation will be more efficient.