LEAD-COOLED FAST-NEUTRON REACTOR BREST Yu.G. Dragunov, V.V. Lemekhov, A.V. Moiseyev, V.S. Smirnov (NIKIET, Moscow, Russia) Large-scale nuclear power based on fast-neutron reactors operating in a closed nuclear fuel cycle (CNFC) has a potential for stopping the growth in the consumption of fossil fuels, assume most of the share in the increase of electricity generation and contribute to the accomplishment of the task voiced by Russian President V.V. Putin at the Millennium Summit at the UN headquarters regarding the sustainable development of humankind having in mind the radical solution to the problem of nuclear nonproliferation and global environmental improvement. However, nuclear power of such a scale will prove to be socially acceptable only when, while remaining economically competitive to alternative energy sources, it naturally excludes the potentiality for the most severe accidents with radioactivity and toxicity release requiring evacuation of population. A new approach to the selection of fundamental designs is suggested for overcoming the contradictions between the safety and cost-effectiveness requirements in the development of a new-generation fast-neutron reactor. It consists in implementing consistently the concept of natural (or intrinsic) safety achievable primarily thanks to the natural dependencies, physical and chemical properties and qualities inherent in the neutron balance in the fast reactor, in the fuel and in other reactor components rather than through building up expensive engineered barriers and complex safety systems. Also important are engineering solutions that contribute to the above properties to be implemented in full. 1. Design features of the BREST reactor An integral review into the innovative reactor technologies of a new generation under consideration in Russia and elsewhere shows that the concept of a fast-neutron reactor with a heavy liquid-metal coolant meets higher safety and fuel supply requirements. It is exactly this quality that was used as the basis in respective pioneer designs in Russia and later adopted in Europe (ELCY, ALFRED-FALCON, MYRRHA), as well as in the USA, Japan, China and South Korea. The recognition of the fact that heavy metals are prospective as coolant materials has been reflected in quite an active international cooperation, primarily as part of GENERATION IV and IAEA programs [1, 2]. Under development in Russia is BREST-OD-300, an intrinsically safe pilot demonstration lead-cooled reactor with uranium-plutonium nitride fuel. The BREST-OD-300 reactor is also viewed as a prototype of future commercial BREST-type reactors for largescale naturally safe nuclear power. Therefore, the selection of the BREST-OD-300 s basic designs and performance, including the power level of 700 MW(th), a two-circuit configuration of the heat removal system with subcritrical water-steam used as the secondary circuit fluid, as well as of other designs, is dictated not just by the intent to demonstrate the natural safety properties of this reactor technology but also by the requirements for providing the continuity of the fundamental designs in future BREST-type concepts of a higher power. The development and construction of the BREST-OD-300 reactor is one of the tasks under the energy strategy of Russia for the period up to the year 2030 approved by the Russian Federation government order No. 1715-r, dated 13.11.2009, Development Strategy of Russian Nuclear Power in the First Half of the 21 st Century approved by the Russian government on 25.05.2000, the Federal Target Program Nuclear Power Technologies of a New Generation for the Period of 2010-2015 and up to the Year 2020 approved by the 1
Russian government in 2010, and the Proryv project (2011) that unites projects for the strategic achievement of targets in the formation of naturally safe nuclear power technologies based on fast-neutron reactors and a closed nuclear fuel cycle. 1.1. Combination of the fast reactor, lead coolant and nitride fuel properties as a natural safety factor Primarily, the selection of the lead coolant and a dense heat-conducting nitride fuel has been dictated by the natural safety requirement. The use in the BREST s integral design of a high-boiling (~2000 K), radiationresistant and low-activated lead coolant, which is inert when contacting water and air, does not require high pressure in the circuit, and excludes accidents with a loss of coolant and heat removal, and with fires and explosions from contacts with the environment (water and air). Using a dense (a nuclear concentration of heavy metal is 20 % as high as in traditional oxide fuel), heat-conducting (~ 5 to 7 times as high as in traditional oxide fuel) nitride fuel compatible with the lead coolant and the fuel cladding steel enables operations with relatively low working fuel temperatures (T max <1300 C), a small thermal energy inventory, a minor release of gaseous fission products from the fuel and their low pressure on the fuel cladding, which contributes to its integrity being preserved. Combining the properties of a heavy lead coolant and a high-temperature dense uranium-plutonium nitride fuel creates the conditions for achieving the complete plutonium breeding (CBR 1), which, along with a small fuel temperature power effect, enables power operations with a small operating reactivity effect of below the effective delayed neutron fraction (β eff ), and excludes catastrophic consequences of an uncontrolled power growth with the complete reactivity margin due to equipment failures and personnel errors being implemented. Minor moderation and absorption of neutrons by lead makes it possible, without worsening the reactor s physical characteristics, to expand the fuel element lattice, increase the coolant flow path and, thanks to this, raise the level of power removed when the reactor is cooled down by the natural circulation of lead. A large volume (~1000 m 3 ) of the lead circuit that accumulates the generated heat ensures that emergency processes and transients run smoothly with no notable increase in the circuit temperature. The presence of a passive coolant flow rate reduction feedback prevents the temperature limits of safe operation from being exceeded during loss of forced lead circulation. The integral layout of the lead circuit with passive and time-unlimited direct removal of residual heat from the circuit by natural air circulation, heat being discharged into the atmosphere, excludes accidents with fuel and coolant overheating during the reactor cooldown. Therefore, it is only thanks to the natural laws of the chain reaction development in a fast reactor, the properties and qualities of the BREST s major components (lead and coolant), and the designs that make it possible for these to be implemented, that two classes of the most severe accidents have been practically excluded (those with an uncontrolled power growth and a loss of heat removal), and the development of anticipated operational occurrences has been mitigated. 1.2 Lead circuit design and circulation system The BREST-OD-300 has an integral layout of the lead circuit components accommodated in the central and four peripheral steel-lined cavities of the concrete vessel (Fig. 1). 2
The central cavity accommodates the reactor core with a side reflector, the CPS rods, an in-reactor SFA storage and a reactor core barrel that separates the hot and cold lead flows. The four peripheral cavities (one for each loop) accommodate steam generators and reactor coolant pumps, heat exchangers of the emergency and normal cooldown systems, filters and other components. The cavities are hydraulically interconnected. Fig. 1. BREST-OD-300 reactor unit 1. RCP; 2. CPS column on a rotary plug; 3. Steam-water headers; 4. Core barrel; 5. Steam generator; 6. Reactor core; 7. Vessel The lead circulation in the circuit is organized using the head level (Fig. 2). The pumps pump the cold lead to the circuit s upper level from where it flows down by gravity to the level of the core inlet, passes the core upwardly, while being heated, and further goes up to the level of the steam generator inlet where, flowing down, it gives heat to water and steam and then enters the pump chambers. Thanks to the potential energy stored in the level, the selected pattern ensures that the coolant circulation through the reactor core is continued when one or more reactor coolant pumps trip. This pattern also minimizes the potentiality of the steam bubbles going out when a steam generator is depressurized entering the core together with the coolant. Accordingly, there is no positive reactivity effect causing a power growth. 3
3 4 1 2 Fig. 2. Coolant circulation in the BREST-OD-300. 1 core; 2 steam generator; 3 RCP; 4 ECCS channel Routine operation mode 1.3. Features of the reactor core and CPS rod design The BREST reactor core is composed of hexagonal shroudless FAs with fuel rods clad in ferrite-martensite steel. The selection of the shroudless FA design is explained by a number of reasons (including a potential improvement in the progression of anticipated operational occurrences with the flow passage closure at the FA inlet to reduce the share of structural material in the reactor core and with the respective reduction in the fuel load), as well as by a number of other reasons. For the power distribution and coolant heat-ups to be leveled in the radial direction, the reactor core is designed as two radial zones the central zone and the peripheral zone the FAs in which differ only in the diameter of the fuel elements: it is smaller in the central zone and larger in the peripheral zone. Along with a small reactivity margin, the use in all FAs of the fuel of one and the same composition, provided the core breeding ratio is ~1, ensures that the FA power and the coolant heat-ups are stable throughout the lifetime. Apart from the fuel elements, some of the FAs in the central zone include a control and protection system (CPS) rod (Fig. 3) accommodated in the vertical channel of the FA s central portion. The CPS rods, taken together, form two independent reactor shutdown systems, one of which (that composed of scram rods) and the other (composed of shim and automatic control rods) form the second system. The drives of the CPS rods are installed on the upper rotary plug and the rods as such (when in a withdrawn condition) stay beneath the core. For refueling, the drives are disengaged from the rods which are retained in the core under the action of buoyancy force and ensure that the reactor is deeply subcritical. 4
Fig. 3. Reactor core map. 1. Central zone FAs; 2. Peripheral zone FAs; 3. Active CPS rods; 4. Active-passive CPS rods; 5. Shim rods; 6. Automatic control rods; 7. PFBS block; 8. Removable reflector block The reactor core is surrounded by rows of changeable side lead reflector blocks in the form of tight steel hexagonal housings filled with low-rate flowing lead coolant. Some of the side reflector blocks adjoining the core have the form of vertical channels with lead columns plugged at the top and at the bottom (of the gas bell type). The lead column level tracks the coolant head (pressure) and affects the neutron escape. Using these devices (channels of the passive feedback system (PFBS)), the reactor reactivity (and power) is passively linked with the coolant flow rate (head) through the reactor core, which is an important factor of the reactor safety and control. When the reactor operates at the rated power, all CPS rods, except two groups of automatic control rods which are responsible for compensating for the operating reactivity effect (~0.4 ), are withdrawn from the core. Regulating and keeping power in the working eff power range also supports the PFBS by responding to an increase or a reduction in the coolant flow rate when the power changes for keeping it permanently heated up. 2. R&D on the BREST project As part of the FTP and the Proryv project, the detailed design of the BREST reactor is being developed, including calculations and experiments conducted to justify the engineering and process approaches. The respective R&D activities are carried out in a broad cooperation of NIKIET with such nuclear organizations as FEI, ATOMPROEKT, VNIINM, IK ZIOMAR, TsKBM, NPO TsNIITMASh, KBSM, TsNII KM Prometey, VNIITF and others. 5
2.1. Reactor core and safety The series of research and optimization calculations have helped to configure the reactor core so that it met the reactor design performance in terms of ensuring the reactivity margin, the negative integral density reactivity effect and other essential criteria. As the result of analyzing the activities in 2010-2013, key solutions have been found for the base core option. The design of a hexagonal shroudless FA composed of bare grid-spaced fuel elements was developed. The reactor core is encircled by the reflector and shielding block rows. The designs are based on results of precision computational simulation, including of the initial core load based on power-grade plutonium in conditions of a closed nuclear fuel cycle. The fundamental principles of an equilibrium reactor core have been confirmed, including maximum reactivity effect at the rated power (< β eff ), unique neutron field stability, and complete fuel breeding inside the core with power-grade plutonium. A safety analysis has shown that the potential release of the design reactivity margin does not lead to the reactor runaway with an uncontrolled power growth and severe accidents. The reactor transients during normal operation and anticipated operational occurrences have been calculated, updated progression scenarios of anticipated operational occurrences have been developed and the reactor safety has been analyzed. As the result of a deterministic safety analysis for the postulated multiple failures of systems and components, a conclusion was made that the development of initial events for anticipated operational occurrences accompanied by multiple failures of systems and components or human errors, including a complete failure of the two reactor shutdown systems, does not lead to severe accidents (Fig. 4). T, C T fuel T cl T core in. T SG in. T SG out T core in t, c Fig. 4. Transient response plot for blackout with two shutdown systems failing. 6
The plot shows the behavior of the fuel, fuel cladding and lead coolant temperatures at the reactor core and steam generator inlets and outlets It was assumed in the simulation of the refueling and fuel regeneration process that the fuel was cleared of fission products and made up with dump uranium of a predetermined composition after being unloaded from the core and cooled for two years, provided the fuel density was preserved. The accumulated quantity of minor actinides (neptunium, americium) is sent back to the regenerated product for transmutation. To raise the accuracy of calculations, activities are under way to verify neutronic codes, including with the use of a large-size physical bench at FEI and the thermal-hydraulic performance of components, including liquid-metalcooled ones. To demonstrate the serviceability of the developed FA design, an experimental justification program has been prepared under which studies into the fuel cladding and spacer grid corrosion resistance, as well as hydraulic, mechanical and vibration tests of full-scale FA mockups have been scheduled. 2.2. Lead coolant technology Since the results of the earlier activities for the conceptual justification of the lead coolant technology (a new one for power-grade plants) proved it to be sufficiently mature, recent efforts have been focused on the development and testing of individual components and process systems. The tested mockups include those of oxygen detectors, a gas handler (for varied parameters), a hydrogen igniter, a mass-transfer apparatus, and lead and gas filtering materials. As earlier, the key tasks are to develop processes of the circuit filling while keeping the coolant in a reactor-clean condition, as well as processes and technologies for passivating the reactor primary circuit surfaces, depending on the surface treatment. 2.3 R&D to justify component desings A fundamental concept used in the BREST-OD-300 design is that steam generators are accommodated directly inside the primary circuit. This concept may be implemented due to lead being inert to water. A number of justifications for such design are required, both in terms of serviceability and safety. Specifically, experiments have been conducted on test sections with a small number of tubes, the SG thermal-hydraulic availability in the range of operating modes has been demonstrated, and the SG analytic codes have been verified. Absence of a dependent failure in the event of a tube rupture has been confirmed experimentally on a simulator bench for the verification of codes with accidents 7
involving ruptures. Values of the coefficient of heat transfer in lead flowing over transversely the heat-exchange tube bundle have been obtained. Another fundamental approach used is that the emergency cooldown system has its heat exchangers accommodated directly inside the primary circuit. For these, the heat-exchange tube performance in different operating modes, including flow rates at the natural circulation level, have been determined experimentally. Simulations with extended conditions and modes have been planned. A strategy for the parallel simulation of flow passages based on lead and water has been selected for testing the main circulation pump. Characteristics of the model flow paths have been obtained, with the resultant values used to select the optimization fields as applied to the reactor circuit. The use of a metal-concrete reactor vessel dictates the need for the serviceability of its components to be justified, the respective fabrication technology to be tried out, and codes to be verified. A full-scale mockup of the reactor vessel has been built for thermal testing. Concrete samples have been tested, as the result of which the mechanical behavior and variation thereof depending on the operation and drying conditions have been determined. A process mockup of a vessel portion is being built as a follow-up activity. Despite the fact that the reactor shall undergo anticipated operational occurrences with both shutdown system failing, design options with both solidbody and liquid control systems are tried out for the reactivity control systems. Water and air tests have been conducted with lead tests further found to be reasonable to start. It has been found reasonable that the reactor automatics operation algorithms should be conducted not only based on a mathematical model but also using real hardware. A mathematical model of controlled items has been developed and the creation of a control system simulation bench has been planned. 3. International cooperation The development of a fast-neutron lead-alloy-cooled reactor is a part of the GENERATION IV program as one of the six advanced fields of the future nuclear power evolution. The results obtained in the implementation of the BREST project were presented at the meetings and conferences held as part of the Program and during the IAEA-sponsored events and at different international conferences. The results obtained by the other project stakeholders have been perceived. An arrangement has been recently reached with respect to the systematic cooperation and scientific and technical exchange of the BREST project with the European ALFRED- FALCON project as part of the ROSATOM-EUROATOM nuclear safety agreement. The purpose of this cooperation is to improve the scientific and procedural tool, databases, design techniques, and the joint discussion of safety approaches and solutions, that is, to contribute to the progress in the project implementation as the whole. 8
4. Conclusion The BREST-OD-300 innovative fast-neutron natural-safety reactor is developed as a pilot and demonstration prototype of base commercial reactors for the future nuclear power based on new nuclear technologies with a view to the following: practical confirmation of the key designs used in the lead-cooled reactor facility operating in a closed nuclear fuel cycle and of the fundamental guidelines in the natural safety concept on which these designs are based; phased justification of the reactor component service life for the creation of commercial nuclear power plants with lead-cooled reactors. The prototype s power (700 MWth) ensures that there is the design continuity in terms of the reactor core components, as well as in terms of the physical and thermal characteristics for the development of the commercial-power BREST-type reactor facility. The BREST-OD-300 detailed design has been developed based on the experimental research on small- and full-scale equipment mockups and using the existing neutronic, thermal-hydraulic and radiation-physics codes. The calculation results confirm the required level of safety and the implementation of the principles allowed for. In this connection, further follow-up steps will be aimed at justifying the designs using large-scale mockups and experimental facilities, as well as at improving the accuracy of codes and verifying these. References 1. Liquid Metal Coolants for Fast Reactors Cooled by Sodium, Lead, and Lead-Bismuth Eutectic. IAEA Nuclear Energy Series, no. NG-T-1.6, STI/PUB/1567.- Vienna: IAEA, 2012. 2. Annual report 2013 - GEN-IV International Forum. Printed by the OECD Nuclear Energy Agency, 2014. 9