BREST is best1 January 2003
Russia's long-term nuclear power development plans assume use of a new generation of fast reactors that address the major problems of safety, radwaste management, economic efficiency, and the risk of weapons proliferation. Over ten years of study has produced a blueprint for a "naturally safe lead-cooled fast reactor for large-scale nuclear power", known as BREST. By Judith Perera
Professor Evgeny Adamov and professor Victor Orlov of the Research and Development Institute of Power Engineering (NIKIET) have examined the BREST reactor concept in some detail. The preliminary investigations into the physics, thermal engineering, material behaviour and technology of the concept have been completed. These paved the way for design studies for a 300MWe pilot plant and a 1200MWe commercial plant.
The new reactor design had to satisfy the following conditions:
• Unlimited availability of fuel through the use of plutonium from spent fuel, efficient use of natural uranium and, subsequently, thorium.
• Exclusion of severe accidents resulting from equipment failure, human error, or external impact through engineering based on inherent safety.
• Environmentally safe energy production and waste management in a closed fuel cycle.
• Prevention of nuclear weapons proliferation by phasing out reprocessing and enrichment, and through physical protection.
• Improving economic competitiveness and efficiency without adding to the complexity of plant design or imposing extreme requirements on equipment and personnel.
BREST does not "breed" plutonium like earlier fast reactors, but has a core breeding radio of approximately 1. There is no need for uranium blankets, which precludes production of weapons-grade plutonium and it eliminates the need for plutonium extraction through reprocessing.
BREST seeks to address growing public concern about radioactive waste, through a scheme which manages nuclear materials in a closed fuel cycle leaving natural background radiation unchanged. The design maximises the natural safety potential of the reactor by exploiting the physical, thermal and hydraulic characteristics of the core and cooling circuits.
In terms of safety, the reactor design is intended to:
• Rule out the possibility of lead coolant losses with the core voiding.
• Exclude hazardous steam or gas ingress into the core.
• Provide natural lead coolant circulation and heat capacity of the coolant circuit sufficient to overcome transients without exceeding the temperature limits of reactor components.
• Remove residual heat from the coolant circuit by passive cooling systems.
• Withstand seismic loads of magnitude 8 on the MSK-64 scale.
• Have power control in the range of 30-100%.
• Be shut down in case of abnormal coolant temperature rise at the core outlet and reduction of coolant flow in the core by passive threshold-response devices.
• Comply with current regulatory requirements.
BREST is designed as a multi-functional reactor for power production, to use plutonium, to produce radioisotopes for industry and medical applications, and to transmute long-lived fission products generated in reactor operation. The main operating mode is power production in base-load operation, although operation at reduced levels is also possible.
It has a semi-integral arrangement of the lead circuit in a metallic vessel. The vessel, 19m in height, has a diameter of 5.5m at the bottom and 11.5m at the top. The wide upper part of the reactor vessel is separated from its narrower central part by a barrel that forms an annular chamber. In this semi-integral arrangement, the steam generators and the main circulation pumps are placed in the annular chamber, outside the central part of the vessel.
The core and the support structures are secured to the dividing barrel separating the ascending and descending flows of hot and cold lead. It is positioned in the lower part of the reactor vessel. The dividing barrel rests on brackets welded to the pressure chamber barrel forming the inner wall of the annular chamber.
The reactor vessel is a vertical welded tank comprising a top protective cover, upper part (annular chamber) and lower part. It is contained in a steel-lined reinforced concrete vault and rests on 24 roller supports located under the flat bottom of the annular chamber.
The space between the lead coolant level and the top cover is filled with gas. Four pipelines 1m in diameter connect the gas space with emergency pressure relief receptacles to cope with an accidental loss of steam generator (SG) integrity.
To circulate the lead coolant, it is pumped to 2m above the lead level in the inlet chamber and is delivered to the free level of the annular pressure chamber. The coolant goes down the annular gap between the reactor vessel and the dividing barrel to the spacer grid of the core. After being heated in the core to 540°C, it rises to the level of SG nozzles, and enters the intertube space of the SGs. As it goes down, the lead transfers heat to the secondary coolant. Cooled to 420°C, the lead goes up through annular gaps between the load-carrying casings of SGs and their outer dividing barrels and enters the pump inlet chambers and is pumped again to the free level of the reactor pressure chamber.
At normal speed, the axial-flow pump impellers are 3m below the free level in the inlet chamber to provide a pressure-induced cavitation margin. This prevents steam ingress into the core in an accident involving loss of tightness of SG tubes. The coolant comes to the free level twice on its way through the circuit, with the result that steam bubbles rise to the surface and escape to the gas space.
The top protective cover seals the reactor vessel and carries the SGs, pumps, rotating plugs and other devices. The cover is a metal frame filled with concrete and comprises a central ring and an outer ring connected by radial ribs. At the bottom, the cover has a thermal insulation layer of steel sheets interlaid with foil. The cover is cooled by naturally circulating air. From the reactor hall, the air flows down into a slot clearance in the lower part of the plate. Heated air rises and enters the top air cavity connected to the stack by air ducts. The outer diameter of the plate is 11.75m and its height is 2m.
The annular chamber of the vessel is designed to house the SG inlet chambers, pump pressure chambers, check valves of auxiliary pumps, and other structural components. It comprises an outer barrel with a flat ring-shaped bottom. It has a barrel welded to the upper side, separating the MCP suction chamber from the coolant pressure chamber. The bottom is welded to a supporting circular frame. The SG and pump casings have double walls, with the inner casing designed to operate at 5MPa and the outer case at 2.5MPa. Under core cool-down conditions, natural air circulation is set up in the gap between the casings.
The lower part of the reactor vessel is a cylinder with a spherical bottom. The barrel of this cylinder rests on the supporting frame of the vessel, is secured by internal locks and is sealed by welding.
The total mass of the vessel is 1075t and its normal operating temperature is 420°C. The maximum positive pressure is 1.7MPa.
In-pile storage is designed to hold spent fuel assemblies until they cool to <2kW (per fuel assembly) and for interim storage of fresh fuel assemblies. Decrease of heat release prevents fuel rod overheating when spent fuel assemblies are removed from the reactor.
Steam generation follows a once-through pattern with the lead coolant and steam moving in counterflows. The steam generator has a thermal power of 87.5MW and a steam capacity of 186t/h. It is a vertical component with a chamber for feedwater delivery and two chambers for superheated steam removal. Each chamber comprises a cylindrical case with a nozzle, a tube plate, and fasteners. Flat lids with welded membranes seal the chambers. The chambers are connected to the SG lid through cylindrical adapters to equalise temperature distribution and reduce thermal impacts on the lid. There are 374 heat exchange pipes fixed in tube plates running from each chamber. The heat exchange surface is a coiled bundle of pipes, with their ends connected to the tube plates.
The separable chamber lids provide access to the heat exchange tubes, making their inspection and isolation in case of leaks possible. The large bending radius of the heat exchange tubes means they can be examined by nondestructive methods.
Lead coolant is fed into the heat exchange bundle from the top and moves down, giving off heat to the secondary coolant. The SG has two feedwater supply chambers and two similar chambers for superheated steam removal. Horizontal perforated sheets placed above the SG inlet chamber dampen the hydrodynamic impact of lead during any accidental rise following tube failure and steam ingress into the lead.
This SG design is standard, but lack of operational experience with the different parameters, coolant type and design features will require further experiments.
The reactor vault is designed to: house the vessel with its top protective cover on roller supports; prevent radioactive substances escaping from the facility; reduce radiation exposure of operating equipment; accommodate a thermal shield around the reactor vessel; house the system for air cooling; and to remove residual heat.
The reactor vault is made of reinforced concrete lined inside with 10mm thick steel sheets. The reactor vault is 12m in diameter at the top and 7m at the bottom. The concrete walls are 3.0m thick. The concrete temperature should not exceed 100°C, and this can be met by using a thermal shield and air cooling.
The concrete vault relies on insulating expanded-clay concrete for thermal protection. The vault has an air cooling system comprising a set of downflow and upflow pipe
sections. Penetrations of the concrete vault for the pipelines of the primary auxiliary systems are sealed by three-layer stainless steel bellows.
The core consists of three zones of individual fuel assemblies with different fuel-to-coolant ratios. It is surrounded by detachable reflector blocks and by a fixed reflector. The specific features of the core in a lead-cooled fast power reactor depend on the properties of lead that restrict coolant heating, though allowing a larger volumetric fraction of lead compared with sodium coolant. The fuel assembly design is intended to ensure insignificant deterioration of fuel cooling in case of local flow blockage at the inlet or in the fuel portion of the assembly. The assembly must have a small enough hydraulic resistance to guarantee that at least 10% of nominal power will be removed by natural circulation of coolant.
Above the fuel column, there is a 900mm tall plenum to collect fission gas. Coolant temperature gains and maximum cladding temperatures are controlled through the radial three-region shaping of power density and coolant flow rates in the core. The core is divided into central, middle and peripheral radial zones, each with its own fuel assemblies that differ only in the diameter of the fuel elements. Thus, three modifications are used, with rods of a smaller diameter in the central core where lower power density and higher flow rate are required, while at the periphery the fuel rods have a larger diameter. Fuel composition is the same in all fuel rods, so they have roughly similar ratios of fission-to-capture. Therefore the quantity of fissile material remains practically unchanged throughout the lifetime, as do the distributions of power density, temperatures and heat gains.
To prevent thermal and mechanical interaction between claddings and fuel during the irradiation-induced swelling of the latter, there is a 0.2mm gap between the fuel pellets and cladding. The gap is filled with molten lead, which affords low thermal resistance of the fuel rod and reduces the maximum fuel temperature to below 1200K.
The core is surrounded by a lead reflector composed of 90% lead and 10% steel on the side reflector, and composed of 70% lead and 30% steel for the axial reflectors.
The BREST core has no blankets, and the molten lead reflector also functions as coolant. Any discharge or fall in the level of coolant in the reactor introduces negative reactivity. The relatively small size of the core, absence of blankets and small amount reactivity in the core makes it possible to place the working elements of the control and protection system outside the core in the side reflector row nearest to the core. Hydraulic and pneumatic drives are used along with the traditional mechanical drives. Some working elements are designed as columns of liquid lead located in vertical channels with axially varied lead level, which allows varying neutron leakage from the core and reactor.
The core is loaded with shroudless fuel assemblies of square cross-section fixed in supports. The fuel assembly lattice has 121 square cells of which 114 are taken up by fuel rods and seven by guide tubes that, together with spacer grids, form the fuel assembly framework. The column of fuel pellets, made of mixed depleted uranium and plutonium nitrides, has an effective density of 13.5g/cm3.
The fuel assembly design is similar to that of PWR and VVER fuel assemblies, where fuel elements are arranged in a square and hexagonal grid, respectively. With closely packed fuel elements, a hexagonal grid has better thermal characteristics, but with wider spacing these advantages are lost. A square grid was chosen for BREST as fuel elements can be placed at regular intervals at the boundaries between adjacent fuel assemblies.
The assembly includes:
• A stem with fastener, controlled spacer grid and installation encoder.
• A bundle of fuel elements fixed in a frame formed by the end grids and support tubes. The upper part of the grid is designed to act as thermal expansion booster and seismic shock absorber.
• A head comprising elements for positioning and alignment of the refuelling machine.
As coolant temperature at the core outlet reaches 870K, the core radius in the midplane grows by 3 to 3.5mm, leading to negative reactivity comparable with the power effect of reactivity.
The key component in assembly spacing is the seismic shock absorber, which is a hydraulic resonance circuit whose restorative force grows with any increase in disturbance frequency. At frequencies above 1Hz, it maintains pitch and keeps core geometry stable with overloads of up to 1g. The spacer on the stem stabilises the pitch in case of low coolant temperature at the core inlet. The spacer grid material has a smaller linear expansion effect than that of the core support structure material. As a result, the reactivity effect of a 30ºC temperature decrease at the core inlet is practically zero.
A can-type fuel rod consists of high-density U-Pu mononitride fuel pellets, a cladding with plugs, and a heat-conducting contact layer of molten lead. There is a gas plenum in the upper part of the fuel rod to collect fission gas. This is filled during fuel manufacture with helium containing 3% of nitrogen at a pressure of 0.03-0.05MPa.
Fuel pellet diameters, pellet-to-cladding gap, the size of the gas plenum, and the corresponding permissible linear heat rate of the fuel are chosen to keep thermal and mechanical cladding stresses at moderate levels, both in normal full-power operating conditions and in accidents. The height of the fuel pellet column in the core is 1.1m, and the gas plenum is 0.9m high. The minimum radial gap between fuel pellet and cladding is 0.20mm. The density of the pellets is about 96% of the theoretical value. Plutonium content (including actinides) is 13%, that of Pu-239 and Pu-241 isotopes about 10% in total. Oxygen and carbon impurities are limited to 0.15% (weight) for each.
The fuel cladding is a thin-walled tube of 12% chromium ferritic-martensitic steel. It shows high corrosion resistance in lead, small radiation-induced geometry changes, and satisfactory temperature dependence of strength and creep. The outer diameter of the tubes to be used as claddings in the central, middle and peripheral regions of the core is 9.1, 9.6, and 10.4mm, respectively. The tube wall thickness is 0.5mm in the first two and 0.55mm in the third. The initial leaktightness of the fuel element corresponds to initial gas penetrability <10-9l/s, with average gas penetrability of fuel elements in the core 10-8l/s. The maximum linear heat rate of the fuel elements of each type, from the core centre to periphery, is 427, 413 and 353W/cm.
The high inlet temperature of the coolant, dictated by the need to keep a margin to lead freezing point, helps to prevent low-temperature radiation-induced embrittlement of the steel cladding, but restricts the limits of coolant heating and power density in the core.
Choice of fuel
Uranium-plutonium fuel can be used in fast reactors mainly as oxide, metal or carbonitride. In principle, any of these fuels can ensure breeding and economical uranium consumption. However, with oxides, this is attained at the expense of other reactor properties, such as the need to use a uranium blanket. The low heat conductivity of oxides is responsible for high operating temperatures, small melting margins, low confinement of gaseous and volatile fission products, and a large power-related reactivity effect.
Any of the fuels can reach high burnups, but metallic fuel has to be heavily doped and made sufficiently porous, negating any advantages over carbonitride on density, heat conduction and full reproduction in the core. Carbonitride fuel has less swelling and is much better in holding fission gas at moderate temperatures, which, given a small clearance between fuel and cladding and a pocket to collect fission gas, results in considerable reduction of cladding stresses caused by fuel swelling and gas pressure.
Carbide and nitride fuels interact only slightly with steel claddings. Mononitride fuel has advantages over carbide fuel in terms of density, swelling and fission gas containment, and, with moderate heat rates, has a considerable disintegration temperature margin.
However, the main reasons for preferring mononitride to carbide fuel were the high oxidation rate and ignitability of carbide, which cause problems during fabrication, and irradiated fuel handling in case of cladding failure. The shortcoming of mononitride fuel is perceptible neutron absorption in the reaction N-14 (+n, -p), which leads to neutron balance impairment and formation of environmentally hazardous C-14. This problem is not very important for a pilot reactor of medium capacity or for a small series of such reactors. For a large-scale power industry, it may prove preferable and more economic to use nitride fuel enriched with N-15.
Preliminary calculations show that a reactor with mononitride fuel should have a minimum capacity of 300MWe. It was demonstrated that replacing traditional uranium blankets with a lead reflector to exclude weapons-grade plutonium production made it possible to: reduce neutron leakage and lead density coefficient of reactivity; make the reactivity effect of lead level strongly negative; equalise radial and axial distribution of power density; and increase temperature margins and average fuel burnup.
Fresh and spent fuel
The estimated corrosion and radiation resistance of fuel claddings determine fuel lifetime in a lead-cooled reactor. For BREST, the lifetime is 1500 effective days (five calendar years). The interval between regular refuelling operations, with one-fifth of the core inventory in each region replaced, is 300 effective days, and the duration of the external on-site fuel cycle is estimated at two years.
The first core inventory uses plutonium separated from LWR spent fuel during reprocessing and subsequently cooled for 20 years to allow partial decay of Pu-241 into Am. This is then mixed with regenerated, natural or depleted uranium. With operation under steady-state conditions of partial refuelling, several recycles will bring the fuel close to an equilibrium composition, when the mass and isotope content of the loaded and unloaded plutonium and minor actinides will coincide.
Planned core refuelling takes place once a year, with 37 fuel assemblies unloaded in this process. In-pile reloading is carried out using two rotating plugs and an internal reloading machine (IRM) that stands on the smaller rotating plug. The IRM can deal with at least eight reloading operations in a six-hour shift.
Spent fuel assemblies are removed by the IRM from the core, pulled into the housing and carried to the external spent fuel storage facility. If found to be faulty, a spent fuel assembly is placed in a special leaktight canister and transported to external storage. The IRM housing is joined to the storage cover and the machine lowers the spent fuel assembly into a "dry" cooled pocket. The protective cover serves as a biological shield. The storage has room for spent fuel assemblies unloaded over two years and for a whole core loading set for emergency use. Spent fuel assemblies are sent from the storage to fuel reprocessing facilities.
Loading of fresh fuel assemblies involves their removal from casks; head-to-tail heating in an argon environment; and installation in the in-pile storage.