Overview of Fast Reactors Technology and IAEA Activities in Support of its Development

Transcription

1 EC Collaborative Project SILER: Training Course on Seismic Protection of Lead-Cooled Reactors Verona, May 21-25, 2012 Overview of Fast Reactors Technology and Activities in Support of its Development Stefano Monti Team Leader - Fast Reactor Technology Development Division of Nuclear Power International Atomic Energy Agency

2 Organizational Structure

3 Outline Why Fast Reactors FRs: technologies, history, current status and future prospective Key Elements for future Development and Deployment General Framework of Safety of Fast Reactors FR activities and deliverables and the role of the TWG-FR 3

4 Why Fast Reactors Almost all nuclear reactors under operation are thermal reactor, in which a moderator slows the energy of neutrons down to the thermal energies In thermal reactors the amount of fissile material consumed is greater than the one generated from fertile nuclides Only a small fraction of the energy potential of natural uranium is exploited Gas Cooled Reactors 6% Boiling Water Reactors 21% Pressurized Heavy Water Reactors 9% Light Water Graphite Reactors 4% Fast Breeder Reactors 1% Pressurized Water Reactors 59% Reactor Types in Use Worldwide, Jenuary 2004 Uranium resources are under potential stress if only U-LWRs are deployed. The deployment of FRs with high breeding gain and short doubling times (CDT) can help. 4

5 Why Fast Reactors: Conversion Ratio vs. Breeding Ratio Condition for breeding η neutrons are generated by fission per neutron absorbed in a fissile isotope of these neutrons 1 is needed to maintain the chain reaction some others (p) are lost by radiative capture or by leakage Then the number of fissile nuclei produced per fissile nucleus destroyed is: B =η 1 p If B<1 the reactor simply converts isotopes and B is called Conversion Ratio If B>1 the reactor breeds fissile nuclides and B is called Breeding Ratio 5

6 Why Fast Reactors: Favorable neutron economy with respect to thermal neutron spectrum reactors η is significantly higher in the case of fission induced by fast neutron than thermal neutrons Neutron yields for various fissile atoms (Source: A. Waltar, A. Reynolds) Reactor types Natural Uranium Uranium 235 Uranium 233 Plutonium 239 Thermal Fast < Neutron yields in thermal and fast spectrum reactors 6

7 Why Fast Reactors: the ratio fission/absorption is also higher, and actinides are preferentially fissioned Fissile isotopes are likely to fission in both thermal and fast spectrum However, the fission fraction is higher in fast spectrum Moreover, significant (up to 50%) fission of fertile isotopes in a fast spectrum 7

8 Why Fast Reactors: Great flexibility thanks to excess of neutrons and transmutation performances As first discovered by Enrico Fermi in 1944, the nuclear characteristics of transuranics (TRU) cross sections in a fast neutron spectrum allow a great FR flexibility: Breed (i.e. Conversion Ratio CR>1) Sustainability Burn (TRU or MA), i.e. CR<1 Transmutation for waste management Breed (e.g. Pu) and burn (MA) CR~1: Self-sustaining. Wide coolant and fuel type choice according to the objective, e.g. short Doubling Time DT: Na and dense (e.g. metal) fuels Wide range of MA content and different Pu vectors or TRU compositions Few credible competitors 8

9 Identified resources Total conventional resources Total conventional resources and phosphates Why Fast Reactors: Resource preservation The high breeding ratio, combined with the multi-recycling of the spent fuel, allows fast reactors to fully utilize the energy potential of natural uranium, assuring a potential energy supply for thousand of years FR with CR>> * Pure fast reactor fuel cycle with recycling of U and all actinides Pure fast reactor fuel cycle with Pu recycling Current fuel cycle (LWR, once-through) Years * - based on 22 Mtu of phosphates, of identified resources, and of undiscovered resources 9

10 Composition of Spent Nuclear Fuel (Standard PWR 33GWd/t, 10 yr. Cooling time) Uranium, 95.5% Stable Fission Products, 3.1% Minor Actinides, 0.1% Plutonium, 0.9% Short-lived Cs and Sr, 0.2% Longlived I and Tc, 0.1% Other long Lived Fission Products, 0.1% Most of the hazard stems from Pu, MA and some LLFP when released into the environment, and their disposal requires isolation in stable deep geological formations. Decay heat in a repository also determined by Pu and MA 1 tonne of SNF contains: kg U 8,5 kg Pu Minor Actinides (MAs) 0,5 kg 237 Np 0,6 kg Am 0,02 kg Cm Long-Lived fission Products (LLFPs) 0,2 kg 129 I 0,8 kg 99 Tc 0,7 kg 93 Zr 0,3 kg 135 Cs Short-Lived fission products (SLFPs) 1 kg 137 Cs 0,7 kg 90 Sr Stable Isotopes 10,1 kg Lanthanides 21,8 kg other stable 10

11 Why Fast Reactors: Actinides (Pu, Am, Np, Cm) Burning Recycle of all actinides in fast reactors provides a significant reduction in amount, heat load and time required for radiotoxicity to decrease to that of the natural uranium ore used for the LWR fuel (from 250,000 years down to about 400 years) Radioactive waste decay time Storage space requirements 11

12 Why Accelerator Driven Systems Fast neutron systems should be privileged for transmutation. In case of Minor Actinide dominated fuels and no U ( U-free fuels ), the effective delayed neutron fraction can be unacceptably low, with consequences on the safety case, if the core is critical. A way out: a sub-critical, source driven reactor (ADS). However, there is the choice between critical (CR<1) or subcritical cores (ADS). 12

13 Fast Neutron Systems in a Closed Fuel Cycle: Towards a more sustainable nuclear energy Today s generation of reactors: Safe, reliable and competitive Availability of secure resources (about 100 y at the present rate of consumpition) Reprocessing of spent fuel for enhanced use of resources Technical solution for waste management Tomorrow s generation of reactors / Fast Neutron Reactors: The above and. Multiplication by a factor 50 to 100 the energy produced by a given amount of uranium, Minimization of volume, thermal load and radiotoxicity of waste Closing the fuel cycle : enough resources for thousands of years

14 Why Fast Reactors. are not deployed as LWRs? There is the the possibility of designing nuclear reactors without neutron moderator, in which the chain reaction is sustained by high energy neutrons (fast neutrons) Fast reactors can convert Uranium-238 into Plutonium-239 at a rate faster than they consume the fissile nuclides, allowing a full utilization of natural Uranium which is composed mainly by fertile nuclides (the amount of U235 in natural Uranium is only the 0.7%) This potentiality of fast reactors has been recognized since the beginning of nuclear power era, but the intrinsic characteristics of fast spectrum require more complex and expensive technologies (both for the reactor and the associated fuel cycle) which have prevented so far the same successful development/deployment of light water reactors 14

15 Fast Reactors: Technologies, History, Currents Status and Future Developments 15

16 Three main technology options depending on primary coolant Sodium Fast reactor Lead Fast Reactor Gas Fast Reactor

17 Sodium Fast Reactor (SFR) outlook Strong national programs and experience (400 reactor.year) Gather fresh operating experience from existing, new and restarting reactors 4 Key technical issues Advanced fuels including actinide recycling Converge safety approach (main issue: Na chemical reactivity) Resolve feasibility issues regarding in-service inspection and repair Energy conversion systems Codes and standards for high temperature application (550 C)

18 Sodium properties: several advantages Low melting point (97.8 C) Large range of the liquid phase (99 C C) Low density and viscosity Very high thermal conductivity Excellent electrical conductivity Low saturation vapor pressure Transparent to neutron Low activation No specific toxicity Cheap and largely available Perfectly compatible with steels Very limited amount of particles in sodium Low oxygen and hydrogen solubility Very good wetting

19 Sodium properties: Three main drawbacks Very important reactivity with water possible deleterious effects in Steam Generator Units (SGU), in case of pipe rupture Must be avoided or mitigated by design Selection of a modular SGU Must be detected, Thanks to the production of hydrogen Risk of hydrogen explosion has to be mitigated Important chemical reactivity with air Can induce Na fire Need inert zones and confinement Need early detection Opacity Need specific equipments for under -sodium viewing and measurements

20 Lead-cooled Fast Reactor (LFR) outlook Limited experience (LBE-cooled ALFA-class submarines in Russian Federation) Resolve feasibility with respect to components and corrosion/erosion control Key technical issues: Materials Design features High temperature applications: Operating parameters (800 C)

21 Gas-cooled Fast Reactor (GFR) outlook No experience Some benefit from VHTR Key technical issues: SiC clad carbide fuel Safety Components and materials (850 C)

22 Two main LMFR lay-outs

23 Historical records EBR : First SFR and first electricity production from nuclear

24 SFR Experimental Reactors Thermal Power (MW) First Criticality Shut down Country Comments CEFR China RAPSODIE 24 / France DFR (NaK) GB KNK 1 - KNK Germany FBTR India PEC 120 Italy JOYO Japan BR Russia BR 2 0, Russia BR 5 - BR oct Russia BOR Russia CLEMENTINE (Hg) 0, USA EBR 1 1, USA First FR electricity LAMPRE USA EBR USA SEFOR USA FFTF USA

25 SFR - Demonstrators Electrical Power (MW) First Criticality Shut down Country Comments FERMI (EFBR) USA BN (*) Kazakhstan PHENIX France PFR GB SNR Germany Construction stopped 1992 MONJU Japan CLINCH RIVER (CRBR) 350 USA Construction stopped KALIMER Korea Project ASTRID France Project

26 SFR Power Reactors Electrical Power (MW) First Criticality Shut down Country Comments BN Russia SUPERPHENIX France BN Russia SPX France Project stopped SNR Germany Project stopped CDFR 1300 GB Project stopped EFR 1500 EU Project stopped PRISM 1245 USA Project stopped PFBR India Construction JSFR ? Japan Project KALIMER ? Korea Project BN-S 1200 Russia Project CFR China Project

27 SFR in operation: BN-600 in Russian Federation On BN the largest operating sodium-cooled fast reactor in the world - celebrated the 31 st anniversary since it was connected to the grid General parameters Thermal power, MWth Electric power, MWe Design lifetime, year Primary circuit Arrangement Reactor vessel support Primary and secondary coolant Number of heat removal loops Inlet/outlet core sodium temperature, С Sodium flow rate, t/h Core and fuel Fuel Max. fuel burnup, % h.a. Diameter, mm Height, mm Pool-type At the bottom Sodium 3 377/ Uranium dioxide pellets In critical condition more than hours; Electricity production: about 120 billion kwh; Average value of the load factor equal to 73.95%.

28 SFR in operation (suspended): MONJU in Japan Electricity Output : 280MWe (714MWt), Sodium Coolant, MOX Fuel Core Temperature, flowrate etc. Primary sodium reactor vessel inlet/outlet: 529/397 C, 5100 t/h/loop Secondary sodium IHX inlet/outlet: 325/505 C, 3700 t/h/loop Steam at the turbine inlet: 483 C, 12.5MPa, 1137 t/h Secondary sodium R/V loop A loop B SH ACS IHX EV IHX SH ACS TB EV Primary sodium Primary Circulating pump Air cooler (AC) Super heater (SH) loop C Turbine SH ACS IHX EV Generator Secondary Circulating pump Condenser Evaporator (EV) Sea water pump Core Intermediate Heat Exchanger (IHX) Feed water pump Primary sodium loop Secondary sodium loop Water/steam system

29 SFR in operation (suspended): MONJU in Japan Construction start October 1985 Completion of equipment installation May 1991 SST beginning December year 5 month SST suspension Core confirmation test May-July % power confirmation test Power rising test 40% 75% 100% Construction Function test System start-up test (SST) Commercial operation First criticality April 1994 First connection to grid August 1995 Sodium leak accident December 1995 We are here. May 2011

30 SFR in operation (suspended): JOYO in Japan Experimental fast reactor at Japan Atomic Energy Agency s O-arai Research and Development Center 100 MWth sodium cooled fast reactor. First criticality was achieved in April 1977 As materials testing reactors, it has shown excellent performance for more than 26 years. Role of the JOYO experimental fast reactor: Advancement of technology through operation and experiment. Conducting irradiation tests on fuel and materials. Validation of innovative technology for development of future FBR.

31 SFR in operation: FBTR in India 40 MWt (13.5 MWe) loop-type experimental fast reactor located in Kalpakkam. First criticality on 18 October Important works including PFBR shielding experiments, testing of transfer arm in air, boron enrichment, post-irradiation examination of FBTR fuel after 125 GWd/t burnup, structural integrity testing, and reprocessing of carbide fuel are being carried out. Construction, commissioning and operation of FBTR have given considerable amount of experience and confidence FBTR will continue to be the workhorse for the testing of metallic fuels and advanced structural materials being developed at IGCAR for the next generation of fast reactors.

32 SFR in operation: CEFR in China Commissioning of Phase A ended Reactor block installation finished Main building finished Preparation of Site Thermal power, MW 65 Electric power, net, MW 20 Reactor core, Height, cm 45.0 Diameter equivalent, cm 60.0 Fuel (first loading, (Pu, U)O 2, [UO 2 ] Pu, total, kg Pu, kg U- (enrichment), (first loading) kg (%) 42.6 (19.6%), [236.7 (64.4%)] Linear power max, W/cm 430 Neutron flux, n/cm 2 s

33 SFR in operation: BOR-60 in Russian Federation Overall plan: Thermal power, MW Up to 60 Electrical power, MW 12 Primary circuit: Coolant Sodium Coolant temperature, C Core inlet Up to 360 Core outlet Up to 530 Coolant flowrate through reactor, m 3 /h Up to 1200 Reactor core: Maximum neutron flux density, n cm -2 s Maximum core power density, kw/l 1100 Average neutron energy, MeV 0.45 Fuel UO 2 or UO 2 -PuO 2 Enrichment with 235 U, % Maximum contents of Pu, % Up to 40 Enrichment with 239 Pu, % Up to 70 BOR-60 is used for: Material tests; Isotopes production (nickel-63, strontium-89, gadolinium-153 ); Tests of various equipments of fast reactors; Heat and electricity production. In operation since more than 41 years. In December 2009, Rostechnadzor has given the license to the RIAR for further operation of BOR-60 up to

34 SFR under Construction: 500 MWe PFBR in India Full name: Prototype Fast Breed Reactor 500 MWe Designer: BAHVINI, India Reactor type: Fast Breeder Reactor Coolant: Liquid sodium Neutron Spectrum: Fast Neutrons Thermal/Electrical Capacity: 1900 MWt/660 MWe Fuel Cycle: closed fuel cycle Salient Features: Passive decay heat removal system; it does not require water for emergency cooling in the case of an accident; core catcher Design status: 1 unit under construction in Kalpakkam - India; commissioning in

35 SFR under Construction: BN-800 in Russian Federation Reactor thermal power, MW 2100 Unit electrical power, MW 880 Unit net efficiency, % 40 Operation life, year 40 Breeding ratio 1.04 Bird s-eye view of the reactor compartment of the main building

36 Fast Reactor Under Development Large-size SFR KALIMER, Korea Republic of, 600 MWe BN-1200, Russian Federation, 1200 MWe JSFR, Japan, 1500 MWe CFR, China, 1000 MWe

37 13725 DEMO and PROTO Fast Reactors under Development ASTRID, France, 600 MWe ELSY, EU, 600 MWe No Component Main vessel Core support structure Grid Plate Weight in t PFBR CFBR Core Inner vessel ALLEGRO, France, 100 MWth Ø Transfer Arm Large Rotatable Plug SRP/Control Plug IHX CFBR, India, 500 MWe Primary Pump Anchor safety vessel Thermal Insulation

38 Small Modular Fast Reactors under Development PRISM, USA, 155 MWe 4S, Japan, 10 MWe SVBR, Russian Federation, 100 MWe BREST, Russian Federation, 300 MWe HYPERION, USA, 25 MWe SSTAR, USA, MWe TerraPower TWR, USA, 500 MWe

39 DEMO and PROTO Fast Reactors under Development ELSY Name: European Lead-Cooled System - ELSY Designer: Project developed in the framework of EC Reactor type: Pool type Coolant: Pure Lead Plant Size: 600 MWe Design Goals and Characteristic Features: MOX fuel, Spiral tube steam generators, Removable internals, Diversified DHR systems, Thermal efficiency 40% 39

40 SM-FR under Development BREST-OD-300 Name: Bystriy Reactor Estestvennoy Bezopasnosti (Fast Reactor Natural Safety) - BREST Designer: RDIPE, Russian Federation Reactor type: Liquid Metal Cooled Reactor Coolant: Lead Plant Size (Thermal capacity): 700 MWt Plant Size (Electrical Capacity): 300 MWe Design Goals and Characteristic Features: PuN-UN Fuel Material, Fuel Cycle 10 Months, High level of inherent safety due to natural properties of fuel, core and cooling design. 40

41 SM-FR under Development SVBR-100 Name: Svinetc-Vismuth Bystriy Reactor (Lead-Bismuth Fast Reactor) / SVBR-100 Designer: AKME, Russian Federation Reactor type: Liquid Metal Cooled Reactor Coolant: Lead-Bismuth Plant Size (thermal Capacity) : 280 MWt Plant Size (Elctrical Capacity): 106 MWe Design Goals and Salient Features: UO 2 Fuel Material, Fuel Enrichment 16.5%, Fuel Cycle 8 years, Closed Fuel Cycle with MOX, operation in a self-sufficient mode 41

42 SM-FR under Development SSTAR (ANL, USA) CLOSURE HEAD CO2 OUTLET NOZZLE (1 OF 8) CO 2 INLET NOZZLE (1 OF 4) Pb-TO-CO 2 HEAT EXCHANGER (1 OF 4) FLOW SHROUD RADIAL REFLECTOR ACTIVE CORE AND FISSION GAS PLENUM FLOW DISTRIBUTOR HEAD CONTROL ROD DRIVES CONTROL ROD GUIDE TUBES AND DRIVELINES THERMAL BAFFLE GUARD VESSEL REACTOR VESSEL Name: small, sealed, transportable, autonomous reactor SSTAR Designer: ANL, USA Reactor type: Pool type Coolant: Lead Plant Size: 19.8 MWe Design Goals and Characteristic Features: Transuranic nitride (TRUN) Fuel, Core Life Time 30 years, Supercritical CO2 Brayton Cycle, Net Plant Efficiency 44% 42

43 SM-FR under Development Hyperion Name: Hyperion Power Module Designer: Hyperion Power Generation, Inc., USA Reactor type: Liquid Metal Cooled Reactor Coolant: Liquid Metal (Pb-Bi) Plant Size: 70 MWt / 25 MWe Design Goals and Salient Features: Uranium Nitride Fuel Material, Enrichment 19.75%, Fuel Cycle 10 years, Transportable Factory Fueled Design 43

44 MBIR (Russia) Lay-out of the MBIR reactor vessel and its experimental channels Na-cooled Research Fast Reactor aimed at in-pile tests of new types of fuel, structural materials and various FR coolants (Na, Pb, Pb-Bi, etc.) Start-up of MBIR is scheduled in

45 BASIC CHARACTERISTICS OF THE MBIR REACTOR Parameter Value Thermal power, MW ~150 Electric power, MW ~40 Maximum neutron flux density, n cm -2 s-1 ~ Driven fuel Vi-pack-MOX, (PuN+UN) Test fuel Innovative fuels, MA fuels and targets Core height, mm 600 Maximum core power density, kw/l 1100 Maximum neutron fluence per year, n cm -2 ~ (up to 45 dpa) Design lifetime, year 50 Number of autonomous test loops with different coolants up to 4 Total number of experimental subassemblies and target devices for radioisotope production up to 12 (core) up to 5 (radial shielding) Number of experimental channels up to 3 (core) Number of experimental horizontal channels up to 6 (outside reactor vessel) Number of experimental vertical channels up to 8 (outside reactor vessel) 45

46 MYRRHA by SCK.CEN (Belgium) Spallation Source Multipurpose Flexible Irradiation Facility Fast Neutron Source Lead-Bismuth coolant

47 International Initiatives Generation IV International Forum - GIF Technical maturity around 2030 Steady progress Economic Competitiveness Safety and reliability 4th Generation Nuclear Systems for sustainable energy development Significant progress : Waste minimisation Resource saving Security : non proliferation, physical protection China E.U. Russia Opening to other applications : High temperature heat for industry Hydrogen, drinking water

48 The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was established in 2000 International Initiatives - INPRO To help ensure that nuclear energy is available to contribute to meeting the energy needs of the 21st century in a sustainable manner To bring together technology holders, technology users and other stakeholders to consider jointly the national and international actions required to achieve desired innovations in nuclear reactors and fuel cycles.

49 ESNII - The European Sustainable Nuclear Industrial Initiative SFR Reference (proven) technology SFR Prototype Astrid MWe LFR GFR Supporting infrastructures, research facilities, irradiation facilities & fuel manufacturing and reprocessing facilities Alternative technology LFR demonstrators MYRRHA ALFRED Allegro GFR Demo Test bed of GFR technologies Innovative fuel MA transmutation Coupling to heat applications MA fuel micropilot MOX fuel fab unit

50 Key Elements for Future Development of Fast Reactors Technology 50

51 Key Elements for FR Development Driving force and peculiarity is sustainability: Natural resources preservation Waste minimization but maintaining/improving Safety (also in light of Fukushima accident) Economical competitiveness Proliferation Resistance Identify technological gaps to achieve the objectives in those areas represents a key step for future developments 51

52 Key Elements for FR Development Innovation to ensure that enhanced requirements for performance, safety and costs are met research and technology development Clear path from research and technology development to deployment Limited resources (economic/financial crisis + industrial and governmental focus on existing and near-term deployment reactors): need to optimize human and financial resources collaboration 52

53 Technical Areas for R&D and Innovation Advanced Simulation & Modelling (e.g. multi-physics and multiscale computer codes) with higher level of precision, key for: design optimization (e.g. reduce nominal pick temperatures); drastically reducing uncertainty margins; narrowing down the needs of expensive experimental tests (mock-ups, T/H and safety experiments, etc.) Data and computer code verification, validation, and qualification (V&V&Q) through theoretical and experimental benchmarks, including severe accident analyses Comparative assessments of feasibility, performance, and safety characteristics Technological choices / concept selection 53

54 Technical Areas for R&D and Innovation Advanced structural materials Innovative fuels (including MA-bearing fuels) Core performance Primary and secondary system simplification Compact Heat Exchangers New Power Conversion Systems Coolant technologies and Advanced instrumentation In-service Inspection Safety (reactivity effects, passive systems, seismic behaviour, etc.) & Security 54

55 General Framework of Safety of Fast Reactors 55

56 The Fukushima Accident The Accident which occurred at the Fukushima Dai-ichi NPP on March 2011 has led a renewed global attention and concern on the safety of current and future nuclear energy systems Significant efforts are nowadays devoted to identify the main lessons learned from this event On the basis of the lessons learned, a review of safety design approaches and features of any nuclear system is considered necessary in view of future developments of nuclear power 56

57 Implications on Fast Reactors Even current and future fast neutron reactors have to take into account the main lessons learned from the Fukushima accident In particular, extreme external events which may potentially lead to severe accidental scenarios such as Station Black Out, Loss of Ultimate Heat Sink as well as others have to be considered and analysed The introduction of advanced technical solutions and provisions for the prevention and mitigation of these scenarios (beyond design basis accidents) is a key factor to further increase the safety level Even if it is possible to identify common approaches, the safety characteristics of fast reactors are rather different than that of thermal reactors, and they require therefore appropriate and specific solutions 57

58 Main Safety Characteristics of LMFRs LMFR have a number of favourable safety characteristics with respect to other nuclear systems (in particular L/HWR), i.e.: Easy to operate: No pressure at the primary circuit, High thermal inertia, Control by rod position (no Xe effect, no need of soluble neutron poison). Radioprotection level higher than in LWR; Few effluents; High thermal and electrical conductivity; High thermal efficiency; Large coolant boiling margin; Natural convection. 58

59 Main Safety Characteristics of LMFRs However, LMFR have also characteristics representing design challenges for safe operation, i.e.: High power density: need to provide adequate heat removal under all circumstances; Power variation due to neutron leakage at the core boundaries: core reactivity is very sensitive to core geometry; Core coolant void worth is typically positive; Core is not in the most reactivity configuration, with a core inventory of many critical masses: fuel relocation might significantly increase reactivity, potentially leading to very high power generation (1000s x nominal) Core Disruptive Accident 59

60 Main Safety Characteristics of LMFRs FRs intrinsic safety-related features are very different than that of thermal reactors: Intrinsic characteristics of the fast spectrum (shorter life time of prompt neutron, small number of delayed neutrons, high power density) Different coolants and plant configurations Main concerns are related to accident scenarios (e.g. UTOP, ULOF, USTOP, etc.) which can potentially bring to core disruption events Solutions based on both intrinsic physical mechanisms (negative reactivity feedbacks) and engineered safety systems have to be provided in order to assure the required level of safety. 60

61 Defence-in-Depth: multiple redundant active and passive safety systems Two (in GENIV also three ) redundant and independent shutdown systems: diverse, robust and reliable Multiple coolant pumps Redundancy and diversification for DHRS Multiple barriers to the release of radioactive materials: Cladding on fuel pins Primary coolant system boundary Containment building Negative power and temperature reactivity feedback coeff. 61

62 Inherent safety characteristics: preventing severe consequences from unprotected accidents Based on fundamental phenomena such as thermal expansion, buoyancy-driven flow, and gravity (reliability to be quantified) Mainly address ULOF, ULOHS, and inadvertent withdrawal of CR(s) resulting in a transient overpower accident (UTOP) The focus is on the three main conditions for safe operation of the reactor: Avoid large uncontrolled increases in core power, by means of favorable reactivity feedbacks; Avoid insufficient cooling of the reactor core, by means of natural circulation cooling; Avoid rearrangement of fuel that would lead to energetic events, (solid or molten core compaction) by core or Sub-Assembly design. 62

63 Beyond ULOF, ULOHS, UTOP E.g. UTOP where all control rods are uncontrollably withdrawn from the core Even inherent safety features are unable to prevent temperature increases, coolant boiling, fuel melting and fuel pin failure Probability of occurrence less than 10-6 per reactor.year Mitigation: preservation of the mechanical integrity of the reactor vessel favorable dispersal of the molten fuel to prevent energetic recriticalities and to maintain core coolability. 63

64 Lessons learned from Fukushima accident (1/3) Japanese government report to the - 28 key points grouped in the following 5 Groups: 1. Strengthen preventive measures against a severe accident 2. Enhancement of measures against severe accidents 3. Enhancement of nuclear emergency response 4. Reinforcement of safety infrastructure 5. Raise awareness of safety culture 64

65 Lessons learned from Fukushima accident (2/3) Strengthen measures against earthquakes and tsunamis and, in general, extreme external events (with reconsideration of their magnitude during plant design) Combination of events: in particular combination of hazards and combinations of external hazards with accidents Identification of cliff-edge effects associated to hazards Emergency power supply: diversity to the extent practicable and redundancy for suppressing common cause failure including external events Decay heat removal system: reactor cooling even under loss of all AC power supply utilization of passive heat removal capability for DEC diversity to the extent practicable and redundancy for suppressing common cause failure including external events Ultimate heat sink: diversity of the ultimate heat sinks for the decay heat transfer 65

66 Lessons learned from Fukushima accident (3/3) Improvement of the containment Reduction of potential bypass Independence of confinement barriers Environmental impact of chemically hazardous material (e.g., sodium) Fuel storage systems: adequate heat removal status monitoring even under DEC including the loss of all AC power supplies External hazards: due consideration of loss of all AC power supplies following the extreme external hazards seismic events may be accompanied by subsequent events Means of radiation monitoring: adequate radiation monitoring in DEC Design extension conditions: designs/provisions for Prevention and Mitigation of the severe accident consequences. 66

67 Action Plan on Nuclear Safety 11 March 2011: Great East Japan Earthquake June 2011: Ministerial Conference on Nuclear Safety Declared a number of agreed points that should direct the process of learning and acting upon lessons to strengthen nuclear safety, emergency preparedness and radiation protection Requested that prepares an Action Plan building upon the declarations and recommendations of working groups 67

68 Development of Action Plan June to Sept Draft action plan developed Produced in conjunction with Member States Considered the points of ministerial report, working group reports, fact finding mission to Fukushima, and views of the International Nuclear Safety Group (INSAG) September 2011 Action plan approved by Board of Governors 68

69 Action Plan on Nuclear Safety 12 Actions: Safety Vulnerabilities Peer Reviews EPR Regulatory Bodies Operating Organisations Safety Standards Legal Framework Embarking countries Capacity Building Protection of People & Environment Communication, Research & Development

70 Action Plan: progress so far Nuclear safety action team formed Implementation strategy under development Status website established: Vulnerability assessment method created Safety Standards action plan created Draft review of document identifying gaps in safety standards with respect to Fukushima lessons learned so far, available at 70

71 Action Plan Next Steps March 2012: International experts meeting at - HQ to discuss Fukushima lessons June 2012: Report to Board of Governors September 2012: Progress on the implementation of the Action Plan will be reported to the Board of Governors Next General Conference 2012 lessons learned report by the end of

72 Activities in support of Fast Reactors Development 72

73 Main activities of the Programme on Fast Reactor (1/2) Organize regular Topical Technical Meetings for in-depth information exchange related to development, design, construction and operation of nuclear power plants with Fast Reactors (FR), as well as to R&D on Accelerator Driven Systems (ADS) Organize Large Conferences on different aspects of FR and ADS RTD (e.g. Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities FR09, Kyoto December 2009; FR013 Paris, spring 2013) Establish a forum for broad exchanges on technical requirements for and characteristics of 4 th Generation Fast Reactor Systems, also in collaboration with INPRO 73

74 Main activities of the Programme on Fast Reactor (2/2) Carry out Coordinate Research Projects (CRPs) of common interest to the TWG-FR Member States in the field of FRs and ADS Secure Training and Education in the field of fast neutron system physics, technology and applications Provide support to Nuclear Safety and Security Department for preparation of fast reactor Safety standards / requirements / guides Task #1 of the project Support for Fast Reactor RT&D&D : Support Fast Reactor data retrieval and knowledge preservation activities in MSs 74

75 The Technical Working Group on Fast Reactors Members of the Technical Working Group on Fast Reactors Members of the Technical Working Group on Fast Reactors Full Members Belarus Brazil China France Germany India Italy Japan Kazakhstan Korea, republic of Netherlands Russian Federation Switzerland Ukraine UK USA OECD/NEA European Commission Observers Argentina Belgium Spain Sweden Full Members Observers Participants in the 44 th Annual Meeting of the TWG-FR, Institute of Atomic Energy (CIAE), Beijing, China, May

76 CRPs Recently Completed 76

77 CRP on Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS) Advance efforts towards designing a demonstration facility by providing information exchange and collaborative research framework CRP on Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS) Argentina Belarus Brazil China Poland Russian Federation Belgium France Germany Participants Hungary Italy Japan Netherlands Pakistan Spain Sweden Ukraine USA Spallation Source Improve the present understanding of the coupling of ADS spallation sources with multiplicative sub-critical nuclear system 77 Multipurpose Flexible Irradiation Facility Fast Neutron Source Lead-Bismuth coolant

78 CRP on Analyses of, and Lessons Learned from the Operational Experience with Fast Reactor Equipment and Systems Analyses of and lessons learned from the operational experience with fast reactor equipment and systems Participants France Japan India Korea, Republic of Russian Federation Preserve the feedback from commissioning, operation, and decommissioning experience of experimental and power sodium cooled fast reactors Retrieve, assess, review and archive of all the relevant documentation and information Enable easy access to the information from this feedback Produce lessons-learned, synthesis reports of lessons learned and recommendations from the commissioning, operation, and decommissioning of experimental and power sodium cooled fast reactors

79 CRP on Control Rod Withdrawal and Sodium Natural Circulation Tests Performed During the PHENIX End-of-Life Tests (special session at ICAPP-12) Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHENIX End-of-Life Experiments Participants China India Korea, Republic of Switzerland France Japan Russian Federation USA Experimental benchmark exercises (preparatory analyses, blind calculations, and postexperiment analyses) based on the data obtained during the PHENIX End-of-Life tests V&V of methods and codes currently employed in the field of FR neutronics, thermal hydraulics and plant dynamics to achieve enhanced safety

80 CP on Integrated Approach for the Modelling of Safety Grade Decay Heat Removal System for LMR (Report under preparation) INPRO Collaborative Project: Integrated Approach for the Modelling of Safety Grade Decay Heat Removal System for Liquid Metal Reactors" Participants China India Russian Federation EU/JRC Korea, Republic of Reference Design: 500 MWe pool type PFBR Detailed analysis of a DHR system using different codes and modelling approaches to inter-compare the results obtained (7 case studies for different conditions)

81 CRPs to be Completed and to be Launched 81

82 CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel (special session at NURETH-14) Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel Participants China India Korea, Republic of USA France Japan Russian Federation Validation of CFD methods and turbulence models based on Na thermal stratification measurements performed in MONJU during a reactor turbine trip test conducted in December 1995 in the course of the original start-up experiments Thorough assessment of the calculation versus measured data comparisons

83 CRP on Benchmark Analysis of an EBR-II Shutdown Heat Removal Test (to be launched in 2012) Benchmark Analyses of an EBR-II Shutdown Heat Removal Test Expression of Interest China Italy Japan Netherlands Sweden USA Germany India Korea, republic of Russian Federation Switzerland A comprehensive testing program (45 tests!) conducted between 1984 and 1987 A unique set of whole-plant safety tests that demonstrated the potential for SFR to survive severe accident initiators with no damage Two EBR-II loss of flow tests chosen for this CRP: SHRT-17, the most severe of the loss of flow with scram tests SHRT-45, the most severe of the loss of flow without scram tests

84 New CRP on SFR: Sodium Properties, Sodium Facility Design and Safety Guidelines (to be launched in ) This CRP is proposed by France and it is intended to address the needs of standardization of Na physical and chemical properties, the main rules for designing experimental facilities, good practices and safety guidelines The CRP making available validated data and correlations for Na coolant - will also improve the modelling and simulation capabilities in various fields of SFR technology The outputs of this CRP will contribute to an improvement of the future benchmark exercises and of the design of sodium facilities and their safe operation. 84

85 CRP on Source Term for Radioactivity Release under FR Core Disruptive Accident (CDA) Conditions (to be launched in 2012)?????? Reference design for the safety analysis: 500 MWe pool type PFBR Demonstrate through numerical simulations of FPs transport mechanisms that in future FBRs the radioactivity release to the environment is very low even in the extreme case of CDA Under whole core accident, the fission products and radioactive sodium are the basic source for the radioactivity release

86 Conferences, Workshops & Technical Meetings 86

87 Major Conferences FR09 International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, Kyoto, 7-11 December 2009 AccApp'11 - International Conference on Nuclear Research Applications and Utilization of Accelerators, Oak Ridge, Tennessee, USA, 3-7 April

88 Major Conferences FR13 International Conference on Fast Reactors and Related Fuel Cycles - Paris, 3 7 March 2013 (coorganized with Nuclear Fuel Cycle Section and CEA/SFEN- France) AccApp 13 - International Conference on Nuclear Research Applications and Utilization of Accelerators, Ghent, USA, July

89 Technical Meetings & Workshops TM on Fast Reactor Physics and Technologies, Kalpakkam, November 2011 GIF-/INPRO Workshop on Safety Aspects of Sodium Fast Reactors, Vienna, 30 Nov. 1 Dec TM (in cooperation with NKM Unit) on the s Fast Reactor Data Retrieval and Knowledge Preservation Initiative, Vienna, 6-8 December 2011 TM on Fast Reactors In-service Inspection and Repair: Status and Innovative Solutions, Vienna, December 2011 TM on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors, Vienna, December

90 Technical Meetings & Workshops TM on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Effects, Vienna, February 2012 TM to Identify Innovative Fast Neutron Systems Development Gaps, Vienna, 29 February 2 March 2012 GIF-INPRO Interface Meeting, Vienna, 6 7 March 2012 TM on Impact of Fukushima event on current and future FR designs, Dresden, March JAEA Workshop on Safety of SFR, Tsuruga, Japan, June

91 Technical Meetings & Workshops Fourth Research Coordination Meeting of the CRP on "Benchmark analyses of sodium natural convection in the upper plenum of the MONJU reactor vessel, Tsuruga, Japan, April 2012 First Research Coordination Meeting of the CRP on "Analyses of Fast Reactor Safety Tests Conducted in EBR-II, Argonne, USA, June, th TWG-FR Annual TM, Argonne, June 2012 TM on Construction & Commissioning of SFR, Kalpakkam, India, November

92 Main Deliverables 92

93 Forthcoming TWG-FR Technical Publications (1/2) Status of Fast Reactor Research and Technology Development (850 pages TECDOC! in print): Background and overview Operating experience with SFR Sodium-cooled FR Designs HLM-cooled FR Designs Gas-cooled FR Designs Status of FR core R&D Reactor plant engineering technology development Reactor safety design and analysis National strategies, international initiatives, public acceptance and final remarks 93

94 Technical Reports closed to Publication Liquid metal coolants for Fast Reactors: reactors cooled by sodium, lead and lead-bismuth eutectic (in print) Design Features and Operating Experiences of Experimental Fast Reactors (in print) Proceedings of FR09, Kyoto, December 2009 (in print) 94

95 Technical Reports and NES in Preparation BN-600 Hybrid Core Benchmark Analysis: methods to reduce calculation uncertainties of the LMFR reactivity effects (under final editing) Benchmark analyses on the Natural Circulation Test Performed During the PHENIX End-of-Life Experiments (under final editing) Status Report of Accelerator Driven Systems for waste transmutation and energy production (under final editing) Special issue of Nuclear Engineering & Design Journal devoted to the outcomes of the TM on Physics and Technology of Fast Reactors (papers available and under review) 95

96 Technical Reports and NES in Preparation Final Report of the CRP on Analytical and Experimental Benchmark Analyses of Accelerator Driven System Final Report of the CRP on Lessons Learned from the Operational Experience on Fast Reactors (editing just started) Final Reports of the CRP on Control Rod Withdrawal and Sodium Natural Circulation Tests Performed During the PHENIX End-of-Life Tests (first report under review) Final Report of the CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel 96

97 FR Project WEB-site: 97

98 TWG-FR WEB-site 98

99 Fast Reactor Knowledge Preservation: FR Data Base Efforts to preserve knowledge and past experience gathered in the operation of fast reactors and FR and ADS R&D programmes Fast Reactor and ADS Data Bases jointly managed by NPTDS/FR Project and Nuclear Knowledge Management (NKM) Unit

100 FRs Knowledge Preservation: FR-KOS System NKM Unit developed a Fast Reactor Knowledge Organization System (FR- KOS): IT system to retrieve information stored in an international data base The NPTDS/FR Project collaborates with the NKM Unit to update the system and collect data and info to be uploaded into the system First version released to MSs for testing and stimulating contributions to FR-KOS 100

101 Conclusions Nuclear power has the potential to play a prominent role for the future generations energy supply needs. Almost all nuclear reactors under operation are thermal reactors, which don t allow a complete utilization of natural resources, posing concerns on the future availability of resources Fast reactors enhance sustainability of nuclear energy utilization. The characteristics of the fast spectrum and the closed fuel cycle guarantee a potential energy supply for thousand years and a more suitable waste management Fast reactors technology has been brought to a high level of technical maturity in the last decades by the design and construction of experimental and prototype reactors, and several construction projects are currently on going Several countries are engaged in the development of innovative FR concepts, and therefore important research efforts are worldwide devoted to cover technology gaps and improve safety features, especially in the light of the Fukushima event. In order to promote cooperation, international initiatives have been established in last years (GIF, INPRO, ESNII) The supports Member States Activities by providing a forum for information exchange and implementing collaborative research programmes. The Agency s activities in this field are carried out in the framework of the TWG-FR 101

102 Thanks for Your Attention! Atoms for Peace 102

103 Back 103

104 Large Size Fast Reactor Under Development CFR-1000 Name: CFR-1000 Designer: CIAE, China Reactor type: Pool type Coolant: Sodium Plant Size: 2500 MWt MWe Design Goals and Salient Features: MOX fuel, Breeding Ratio targeted at 1.2, Plant Design Life of 40y, Na-Na-H 2 O loops with 3 circuits of primary and secondary loop CFR-1000 diagram

105 Large Size Fast Reactor Under Development BN-1200 Name: BN-1200 Designer: OKBM, Russian Federation Reactor type: Pool type Coolant: Sodium Plant Size: 2900 MWt 1220 MWe Design Goals and Salient Features: MOX Fuel type, Breeding ratio targeted at 1.2, Gross thermal efficiency of 42%, 4 heat removal loops, Plant Design Life of 60 years

106 MBIR (Russia) Lay-out of the MBIR reactor vessel and its experimental channels Na-cooled Research Fast Reactor aimed at in-pile tests of new types of fuel, structural materials and various FR coolants (Na, Pb, Pb-Bi, etc.) Start-up of MBIR is scheduled in

107 BASIC CHARACTERISTICS OF THE MBIR REACTOR Parameter Value Thermal power, MW ~150 Electric power, MW ~40 Maximum neutron flux density, n cm -2 s-1 ~ Driven fuel Vi-pack-MOX, (PuN+UN) Test fuel Innovative fuels, MA fuels and targets Core height, mm 600 Maximum core power density, kw/l 1100 Maximum neutron fluence per year, n cm -2 ~ (up to 45 dpa) Design lifetime, year 50 Number of autonomous test loops with different coolants up to 4 Total number of experimental subassemblies and target devices for radioisotope production up to 12 (core) up to 5 (radial shielding) Number of experimental channels up to 3 (core) Number of experimental horizontal channels up to 6 (outside reactor vessel) Number of experimental vertical channels up to 8 (outside reactor vessel) 107

108 MYRRHA by SCK.CEN (Belgium) Accelerator (600 MeV 4 ma proton) Reactor subcritical mode (~85 MW th ) Critical mode (~100 MW th )