BN-1800: a next generation fast breeder

The design of the BN-1800 includes different key design features compared to the BN-800 unit currently under construction at Beloyarsk in South Urals and scheduled for completion in 2009.

The design draws on the approximately 130 reactor-years of fast reactor operation in Russia, and especially the BN-600 reactor. The BN-1800 design has enhanced reactor and environmental safety, particularly due to the burning-out of minor actinides, improved proliferation resistance, and considerable improvements in its economics.

MAIN PRINCIPLES AND DIRECTIONS

The development of the BN-1800 power reactor unit employed the maximum amount of proven and scientifically grounded engineering solutions previously implemented in BN-350, BN-600 and BN-800 reactors, as well as using new engineering solutions aimed at improving both the safety and economic efficiency of the plant.

The proven engineering solutions include:

• The choice of sodium as coolant.

• Heat transport scheme of the power unit – sodium is used in the primary and secondary circuits, water/steam is a working medium of the power conversion system. This helps to provide a high level of safety in the reactor plant (RP) and, in particular, means that steam generator (SG) leaks are not events involving radioactive release.

• Integral layout of the primary (radioactive) system in the main and guard vessels.

• Plant economics and safety are improved.

BN-1800 POWER UNIT

Views of the reactor are shown in Figures 1 and 2.

The primary circuit equipment items are located inside the reactor vessel and include: reactor core, six intermediate heat exchangers (IHXs) and three reactor coolant circulating pumps. The circulation path of the primary circuit is of the collector type – all primary coolant pumps have a common suction plenum downstream from the heat exchangers and are connected in parallel to the common discharge header through pressure pipes.

Each of the six secondary circuit loops includes intermediate heat exchanger, steam generator, secondary sodium circulating pump and associated pipelines. Each steam generator consists of two vessels, one of them is an evaporator and steam superheater, while the other provides intermediate superheating of steam discharged from a high pressure cylinder of the turbine.

An air heat exchanger performing the emergency residual heat removal functions during loss of main power supply sources is located in each secondary circuit loop in parallel with the main vessels of the steam generator.

There are discharge devices on the secondary side of each steam generator vessel to evacuate sodium-water products to the first stage emergency discharge tank from which a gaseous component is removed to the second stage tank, followed by its release to the atmosphere through special discharge devices.

The tertiary circuit consists of the steam generating portion of the steam generators, main steam lines, one turbine-generator set, its auxiliary equipment and the support systems, equipment for feedwater deaeration, heating and supply to the steam generators.

REACTOR CORE

Minimisation of the sodium void reactivity effect (SVRE) is attained by increase in the core depth to height (D/H) ratio combined with introduction of a sodium layer and an upper absorber shield above the core.

The core design with single enrichment of fuel has been chosen on the basis of preliminary analysis of several core options and has the following characteristics:

• Use of the fuel with the same composition considerably simplifies the fuel cycle flow chart and minimises a range of fuel elements to a single type.

• Relatively low core power density allows the neutron field radial distribution with the marked rise in its central part.

• The neutron field has a maximum in the region where the control assemblies are located, providing a sufficiently high efficiency with a small location area (relative to the core radius).

• The adopted limited diameter of the control rods location zone enables keeping a moderate diameter of the above-core structure in spite of the increased radius of the core.

• Increased neutron leakage from the central core area decreases the SVRE value and allows the core height to be slightly increased.

• Stability of heat release distribution during the core operation allows for optimal distribution of coolant flowrate in spite of the relatively high value of the heat distribution peaking factor along the core radius.

• The operating period of peripheral fuel assemblies (FAs) is increased according to their lower heating rate in order to increase fuel utilisation efficiency.

Reactor core arrangement

The core consists of 1686 different assemblies arranged with the average pitch of 188mm. A central portion of the core consists of 642 FAs with fuel of the same enrichment, six steel assemblies and 37 cells with control assemblies. Passive emergency protection (scram) assemblies are also provided, which ensure reactor trip independent of the availability of control systems – should the primary coolant flowrate drop below a half of its rated value or the coolant temperature increase at the core outlet. Then there is a lateral shield consisting of one row of steel assemblies and three rows of assemblies with boron carbide, arranged in the radial direction. Behind the lateral shield there is an in-reactor storage, the capacity of which enables location of all spent FAs (with some margin), discharged at one refuelling outage. Behind the in-reactor storage, boron carbide assemblies are arranged in three rows of cells forming an additional replaceable lateral shield (see Figure 3).

Fuel cycle management

A batch-type refuelling strategy is used for the core FAs. The duration of fuel operation in two peripheral rows of FAs is extended to four refuelling intervals compared to three for the rest of the FAs in order to equalise the fuel burnup value across the core. Spent FAs are kept in the in-reactor storage during one refuelling interval and then unloaded from the reactor for reprocessing in the external fuel cycle. Reprocessed fuel from spent FAs is used for new FA fabrication.

Core fuel cycle characteristics have been obtained assuming such patterns of the external fuel cycle that only fission products are extracted from spent fuel, which then is made up with appropriate amount of natural uranium to obtain new fuel. Two years exposure and 0.5% fuel losses are presumed for the external fuel cycle.

STEAM GENERATORS

The steam generator is a once-through vessel-type heat exchanger with supercritical parameters on a steam-water side. It includes an evaporator-superheater (ES) and intermediate steam superheater (ISS), interconnected by pipes. The ES and ISS are connected in parallel on the coolant side. Heating and heated media (sodium coolant, working medium) move in counterflow.

The design of the vessel-type SG enables the reduction of its specific steel intensity by a factor of more than five compared with the section-modular-type steam generator. To provide the steam generator functional capability during the specified lifetime the following solutions have been adopted:

• High nickel alloy with 21% Cr and 32% Ni is used as

structural material for ES heat-exchange tubes.

• X18H9 austenitic steel or high-nickel alloy with 21% Cr and 32% Ni is used as the material of ISS heat-exchange tubes.

• Relative pitch of tubes in ES tube bundle is increased (compared to that in the BN-600 steam generator module) to 2.0-2.2 (meaning that at 16mm tube diameter, the distance between tubes will be 16-19mm compared to 12mm in BN-600 evaporator); relative pitch in the ISS tube bundle is kept the same as in the modules of BN-600 SG ISS.

• Coolant speed in the intertube space of long ES and ISS tube bundles is provided at the level of 2-2.5 m/sec; its further increase is not expedient due to possibility of metal vibration wear in heat exchange tubes.

The combination of these measures considerably increases the time until the appearance of holes in adjacent tubes, which could result in water leakage into sodium through a damaged tube. Also, the relatively short time of transport of interaction products from a leakage place to the SG EPS sensors enables the development of highly effective emergency protection system for the SG. Thus enhanced safety of the SG operation is provided.

REACTOR PLANT SAFETY

At the present stage of the BN-1800 design process there is no detailed data on appropriate control, support and localising safety systems of the power unit, therefore it is impossible to build up real scenarios of beyond design basis accidents sequences and evaluate their consequences. However, some ultimate states of the reactor plant have been determined, which are characterised by partial or full failure of active safety systems and elements. The following initial events have been considered:

• Full loss of off-site and on-site reliable power sources (plant blackout) with control rod drives actuation to trip the reactor operating at 100% power.

• Plant blackout without scram at 100% power operation; emergency residual heat removal system in the initial state.

• Erroneous withdrawal of FA from the core at the refuelling.

• Secondary circuit sodium pipe guillotine-type rupture.

• Sodium-water interaction in steam generator boxes.

The analysis has shown that the safe operation limits are not exceeded in any of the aforementioned accidents.

MAIN ECONOMIC CHARACTERICSTICS

Analysis of economic characteristics has been performed in accordance with the document The Methods of Main Economic Characteristics Analysis for NPP, NCGP, NDHP developed by Atomenergoproekt, giving a costing per kilowatt of around $860. The analysis shows that economic characteristics of BN-1800 are comparable with those of similar designs (BN-800, VVER-1000, VVER-1500 – see Table).

V.M. Poplavsky, A.M.Tsibulia, A.A.Kamaiev, Yu.E.Bagdasarov, I.Yu.Krivitsky, V.I.Matveev, SRC – Institute of Physics and Power Engineering, 1 Bondarenko sq, Kaluga region, 249020, Russia; B.A.Vasiliev,A.D.Budilskiy, Yu.L.Kamanin, N.G. Kuzavkov, A.V.Timofeev, V.I.Shkarin, OKBM, Burnakovsky Projezd 15, Nizhny Novgorod, 603074, Russia; K.L.Suknev, V.N.Ershov, S.V.Popov, S.G.Znamensky , Atomenergoproject, 7/1 Bakuninskaya ulitsa, Moscow 107815, Russia; V.V.Denisov, V.I.Karsonov, OKB “Gidropress”, 21 Ordzhonikidze Street, 142103 Podolsk, Moscow region RF, Russia