Lead-Cooled Fast-Neutron Reactor (BREST)

Transcription

1 Joint Stock Company (JSC) N.A.Dollezhal Research and Development Institute of Power Engineering Lead-Cooled Fast-Neutron Reactor Yu.G.Dragunov, V.V. Lemekhov, A.V. Moiseyev, V.S. Smirnov (BREST) (APPROACHES TO THE CLOSED NFC) INPRO Dialog-Forum, IAEA HQ, Vienna, Austria, May

2 Lead-Cooled Fast-Neutron Reactor (BREST) (APPROACHES TO THE CLOSED NFC) OUTLINE: 1. Preamble: Lead-cooled fast reactors 2. BREST OD-300: Main goals of development, state-of-art 3. BREST OD-300: Natural Safety principles 4. Closed Nuclear Fuel Cycle 5. Back-End of the NFC, Radiation Equivalence Principles 6. Conclusion: prospects, problems, collaboration 2

3 FAST NEUTRON REACTOR WITH HEAVY METAL COOLANT An comprehensive analysis of 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. Namely these features formed the basis of respective pioneer reactor 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 collaboration, primarily as part of IAEA programs (INPRPO and others), GENERATION IV, etc. 3

4 MAIN GOALS OF TEHNOLOGY The exclusion of severe accidents of nuclear power plants (reactivity, loss of cooling, fires, explosions), requiring the evacuation of the population; The closed NFC circuit to fully exploit the energy potential of uranium feedstock; Back-end of NFC: a consistent approach to radiation equivalence of definitively buried RAW with respect to the originally used natural uranium raw materials; Technological strengthening of non-proliferation : lack of separation U and Pu when reprocessing spent fuel, the rejection of U blanket with Pu breeding, and non-involvement of enriched uranium in reactor loading (i.e. no uranium enrichment later); Ensuring the competitiveness of nuclear power in compared with other types of power generation /Dragunov Yu.G., et al, Atomnaya Energiya, v. 113, -1, 2012, pp.58-65/ 4

5 THE REACTOR BREST-OD-300: GOALS AND OBJECTIVES OF CREATION Goal practical confirmation of realization of the Natural Safety concept of the lead-cooled fast reactor, operating in NPP mode with closed NFC. Objectives: Life experience in all stages of the life cycle for commercial power units, built according to the chosen concept Complete fuel breeding (equilibrium mode) for self-sustaining Confirmation of exclusion of accidents caused by reactivity and accidents with loss of coolant, requiring evacuation of the population, in the imposition implemented multiple bounce for internal reasons Gaining experience in NPP operating in the closed nuclear fuel cycle /Dragunov Yu.G., et al, Atomnaya Energiya, v. 113, -1, 2012, pp.58-65/ 5

6 NPP: DESIGN CONCEPT For lack of reactivity margin enough for realization of severe reactivity accident. Integral-type arrangement of the first contour to avoid output of coolant outside the reactor vessel, to eliminate lost of coolant. Using of low-activated coolant with high enough boiling temperature, without rough interaction with water and air in the case of depressurizing of the contour. Realization of full breeding of fuel within the active zone solely, burning of the long-lived actinides. Simplifying of the safety systems due to physical features of used materials and design methods. 6

7 The BREST: key components and technical characteristics Collector of SACR MCP Thermal power, MW 700 Electric power, MW 300 Steam production rate, no less than, t/hour 1480 Coolant of the first contour Gas pressure above the lead level: - exceed, MPa - maximal, MPa Average temperature of the lead coolant on the active zone entry/ exit, С Average temperature of the lead coolant on the steam generator entry/ exit, С Lead 0,003-0,008 0,02 420/ /505 Loop number 4 FA number in the active zone 169 Active zone height, mm 1100 Fuel load, t 20,6 Fuel campaign, years 5 Vessel Core Steam generator Burn-up of unloaded fuel (maximum/ average), % HM. 9,0/5,5 7

8 EXCEPTION TO LOSS OF COOLING Use of multilayer reinforced concrete with bimetallic facing and coolant with high freezing point negligible lost of lead through diffusion of lead into concrete in the case of depressurizing of facing. Absence of lock fittings in the first contour impossible to break circulation. The coolant circulation scheme with the over-fall of free levels a guaranteed prolongation of circulation under lost of power supply. The emergency coolant system with natural circulation, transferring heat directly from the first contour to the final absorber atmospheric air. 8

9 The most severe failures of normal operation 9

10 The most severe failures of normal operation 10

11 The most severe failures of normal operation: unplanned full lost of electric power supply, plus failure of two stop systems at once Т, С Т fuel Т clad Т out AZ Т in SG Т in AZ Т out SG Time, s 11

12 The most severe failures of normal operation 12

13 The most severe failures of normal operation 13

14 The most severe failures of normal operation: unplanned output of the absorber rods Т, С Т clad Т out AZ. (MCP stop due to temperature, than PFBS, than passive safety system-т) Т in SG Т in SG Т in AZ. Time, s 14

15 BREST-OD-300: ACTIVE ZONE 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 15

16 FUEL CYCLE FLOW CHART Fuel operation in the BREST reactor Fuel refabrication Fuel cooling (1 year) Fuel regeneration Makeup by natural or depleted uranium Waste (V.S. Smirnov, International Workshop on Innovative Nuclear Reactors Cooled by Heavy liquid metals: Status and Perspectives, Pisa, April, 2012) 16

17 BREEDING RATIO BR ~1,05 Use of mixed uranium-plutonium nitride fuel with high density and high thermal conductivity and low moderating coolant ensure breeding of the fissionable materials in the active zone. Full reproduction of fission materials in the active zone (BRA ~1,05) allows not to have a reactivity margin to burnout and, accordingly, the margin for overclocking (including cold state), leading to severe accidents requiring evacuation of the population. 17

18 CLOSED FUEL CYCLE The ultimate objectives of the BREST-OD-300 project include demonstration of not only the expected physical and operational characteristics and intrinsic safety of this installation as, but also its capability of operating in a closed cycle mode with an equilibrium fuel system. Equilibrium mode of fuel supply means that the reactor operates with complete reproduction of the fissile nuclides in the reactor core (breeding ratio 1) and fuel is recycling through the extra-reactor facilities components of the closed fuel cycle complex. By this mode, the weights and isotopic compositions of Pu and MА in charged (fresh) and discharged (spent) fuel would be virtually the same, and ultimately, the only one burnt component would be 238U, whose mass would be replenished every time as new fuel is produced. V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear Technologies and Environment, 1'

19 BASIC DIAGRAM OF BREST FUEL REGENERATION Expected initial load: mixed mono-nitride: ~13.2% Pu in U-Pu Plutonium isotopic composition corresponds to Pu, extracted from SNF of VVER after 25 years of cooling. (V.S. Smirnov, International Workshop on Innovative Nuclear Reactors Cooled by Heavy liquid metals: Status and Perspectives, Pisa, April, 2012) 19

20 CLOSED FUEL CYCLE Environmentally safe closing of the fuel cycle would be achieved through utilization of specific fuel recycling and refabrication technologies that only require relatively coarse treatment of spent fuel to remove fission products, adding depleted uranium to the treated fuel mix (U-Pu-minor actinides), nitration and fabrication of new fuel. Irradiation time up to planed average burn-up (cca 8% HA) is 5 year. After discharge from the core, assemblies with spent fuel would be placed in at-reactor storage, cooling for 1 year and then being shipped to reprocessing plant. Spent fuel reprocessing and new assemblies fabrication take the next 1 year. V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear Technologies and Environment, 1'

21 EQUILIBRIUM MODE OF FUEL SUPPLY Equilibrium fuel mode presumes stability of reactivity during fuel burning between refueling (during the cycle), within the effective share of delayed neutrons (βeff) Reactivity margin, βeff ОЗР, β эф Кампания реактора, эф. сут Time, days V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear Technologies and Environment, 1'

22 POWER PRODUCTION COMPLEX IN PLAN NPP with reactor BREST-OD-300, reprocessing plant and long term storage on site. V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear Technologies and Environment, 1'

23 RADIATION EQUIVALENCE PRINCIPLE "The principle of Radiation Equivalence" Balance radioactivity (taking into account the dangers of both the biological impacts and the natural migration) between the hazard of natural uranium used to produce an energy in a closed power system and the hazard of long-lived high-level radioactive waste, elaborated by the reactors. Balance of radioactivity in a closed power system with BRESTtype reactors can be achieved: on the base of in-reactor transmutation of actinides and the extraction and controlled long-term (other of 300 years) storage of other highlevel waste (cooling - to reduce the activity of thousands times) before their final disposal. (A.V.Lopatkin, V.V.Orlov, et al. Fuel cycle of BREST reactors. Solving of RAW and non-proliferation problems, ICONE ) 23

24 Potential biological hazard (ingestion) of irradiated fuel Time after recovery, years (A.V.Lopatkin, V.V.Orlov, et al., N-Novgorod, 2007) 24

25 Assessment of optimal parameters for reprocessing of BREST-OD-300 INF by criteria for the impact on the population and the biota on all potential paths Dose, Zv Radio-equivalence is achieved after 300 years storage provided extraction of Am with residual in the waste is less than 1% Time, year The total dose of external and internal exposure of Man by actinides of BREST-OD-300 INF provided the residual content of U, Pu, Np - 0,1%: - 1 ton of natural U (1); hereinafter: - the residual of Am in LLHL RAW - 100% (2); 10% (3); 3% (4); 1% (5); 0,3% (6) and 0,1% (7). (E.V. Spirin, R.M. Alexakhin, S.I. Spiridonov, The radiation balance of the spent nuclear fuel according to the criteria of impact on human health and the environment. XLIII radio-ecological reading in memorial of V.M. Klechkovsky) 25

26 TECHNOLOGICAL SUPPORT OF NON-PROLIFERATION All FAs of the core contain the same mass of Pu. No uranium blankets and no breeding of weapon-grade Pu, because blankets do not needed. No need to recover Pu for fabrication of reloading fuel (it is suffice to separate fission products and add depleted U). Hence, reprocessing may be used, but it is not suitable for Pu recovery. No need for U enrichment. Surplus Pu is used as part of U-Pu mixture for fabrication of the first loads for new reactors Reprocessed fuel is partially cleaned from fission products (recycling fuel contains about 10-2 FPs present in spent fuel) and incorporates minor actinides, which makes fuel highly radioactive (as radiation barrier to fuel thefts). As projected, reactors burn 238U added into fuel at refabrication. Pu is part of fuel and recycles in the closed cycle as part of highly active fuel (combustion catalyst for 238U ) (A.V.Lopatkin et al., 26

27 COMPETITIVE ABILITY A plant with a BREST-type reactor is expected to be economically competitive owing to the simpler design of the facility and its safety systems, as well as to efficient utilisation of nuclear fuel and generated heat. Low lead pressure in the circuit allows using an integral configuration of the circuit components in a concrete pool, which was tentatively shown to reduce the construction costs. On-site fuel cycle arrangement is also likely to be economically beneficial owing to the shorter out-of-pile cooling and transportation time, which will eventually lead to a reduction in the recycled fuel quantity one of the greater contributors to the fuel cycle costs. BREST-OD-300 being a prototype of the prospective commercial BREST plant, both facilities are quite similar in their design and performance (V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear Technologies and Environment, 1'

28 THE BREST-OD-300 REACTOR DEVELOPMENT The detailed design of the BREST-OD-300 reactor is being developed as part of the PRORYV project, including calculations and experiments conducted to justify the engineering and process approaches. Currently the engineering design of the reactor, including experimental study on small and medium-sized stands and work stations, as well as on the current study on verified software tools. Further justification will be held on the large-scale model. The main areas of research are: the active zone, the main process equipment (reactor vessel, reactor coolant pump, steam generator, etc.), the technology of the coolant, the study of transitional and emergency processes, including the closed NFC issues. 28

29 CONCLUSION 1. Project BREST-OD-300 creates a base for development of commercial reactor for NPP on the basis of new nuclear technologies. 2. Analysis of transient processes in RU BREST-OD-300 shows a possibility of exclusion of heavy accidents, demanding evacuation and displacement of inhabitants while using first physical properties of coolant, fuel, other reactor components, аnd also technical design, directed at it realization. 3. Substantiation of adopted decisions for design project in 2014 is built on experimental substantiation on small- and middle-scale benches and working areas, and also on computational substantiation with verified program means. In the further the substantiation will be held on large-scale mockups. 4. The results of experimental and design works points at possibility of realization in power complexes with RU of a kind BREST in closed fuel cycle basic demands of NP on safety, volume of consumption of fuel raw materials, efficiency, solving problem of spent fuel. 29

30 CONCLUSION PROBLEMS, PROSPECTS, COLLABORATION: Nitride fuel: There are in-pile test of experimental FAs with U-Pu-N fuel in BN- 600 and BOR-60 reactors. Post-irradiation tests (swelling, gas release, etc.) Must be justified: Nitride fuel with MA: MA transmutation mode homogeneous or heterogeneous? Nuclear Safety for active zones fuelled with nitride fuel for the heterogeneous transmutation of MA; Curium - transmutation or storage? 30

31 CONCLUSION PROBLEMS, PROSPECTS, COLLABORATION: The experimental data are needed on the verification impacts of MA on neutron-physical and safety characteristics. Nuclear and reactor data (measurements, verification, libraries). Modeling of MA behavior in radiochemical processes, in fuel, during long-term storage. Risks for proliferation. FP transmutation? Approaches and achieving of the Radiation Equivalence 31

32 ASC «Rosatom» company, JSC «NIKIET» THANK YOU FOR YOUR ATTENTION! P.O.B 788, , Moscow, Russia 32