VO Kernerngie und Umwelt TU Graz. Eileen Langegger

Transcription

1 VO Kernerngie und Umwelt TU Graz Eileen Langegger

2 Generation IV Modul 8 2

3 Energy B people rely on biomass 1 B people no health care due to energy poverty 1.1 B people no access to energy

4 Energy Challenge Population Energy demand Energy security Environment Climate change

5 For 1GW you need

6 Evolution of Nuclear Power Plants

7 Advanced Reactors Evolutionary Reactors Innovative Reactors SMRs ABWR, ACR 1000, AP1000, APWR, Atmea-1, CANDU 6, EPR, ESBWR, VVER 1200 (NuScale PWR, OKBM PWR, B&W mpower, India DAE AHWR, KAERI SMART PWR, OKBM KLT-405 PWR) SCWR GFR MSR LFR Evolutionary Designs (ABWR, ACR 1000, AP1000, APWR, Atmea-1, CANDU 6, EPR, ESBWR, VVER 1200) Innovative Designs (SFR, GFR, LFR, VHTR, SCWR, MSR)

8 Generation IV Participants

9 Generation IV drivers Initiative of US Department of Energy to make nuclear energy part of future energy supply Driver: the 21 st century will see vastly increased demand for electric energy that is sustainable, secure, clean, affordable 10 countries originally 2002: Generation IV Technology Roadmap 2014: Generation IV Technology Roadmap update 3 additional countries 9

10 Gen IV Technology Roadmap Identification of systems that offer significant progress with respect to: Sustainability use of natural resources minimization of quantity and life time of radioactive waste Safety and reliability safe and reliable in operation low probability of core damage eliminate necessity of off-site emergency response Economics low life cycle costs low financial risk Proliferation: unattractive for diversion or theft for use in nuclear weapons Terrorism: improved resistance 10

11 Technology Roadmap 6 design concepts identified: Sodium- cooled Fast Reactor (SFR), Lead Cooled Fast reactor (LFR), Gas Cooled Fast Reactor (GFR), Molten Salt Reactor (MSR) Sustainability Very High Temperature Reactor(VHTR) Safety and reliability Super Critical Water Cooled Reactor(SCWR) Economics 11

12 6 Gen IV Systems 12

13 Goals of Generation IV Sustainability: Generation IV nuclear energy-systems will provide sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production. Generation IV nuclear energy-systems will minimize and manage their nuclear waste and notably reduce the longterm stewardship burden, thereby improving protection for the public health and the environment. Economics: Generation IV nuclear energy systems will have a clear lifecycle cost advantage over other energy sources. Generation IV nuclear energy systems will have a level of financial risk comparable to other energy projects.

14 Goals of Generation IV Safety and Reliability: Generation IV nuclear energy systems operations will excel in safety and reliability, they will have a very low likelihood and degree of reactor core damage and Generation IV nuclear energy systems will eliminate the need for offsite emergency response. Proliferation Resistance and Physical Protection: Generation IV nuclear energy systems will increase the assurance that they are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism.

15 Application of Gen IV Reactors

16 Generation IV General Design Neutron spectrum Coolant Temp. ( 0 C) Fuel Fuel cycle Size(s) (MWe) SFR fast sodium 550 U-238 & MOX closed VHTR thermal helium 1,000 UO 2 prism or pebbles open GFR fast helium 850 U-238 closed 1200 SCWR thermal or fast water 500 UO 2 or MOX open (thermal) or closed (fast) LFR fast Pb or Pb-Bi U-238 closed MSR epithermal fluoride salts UF in salt closed 1000

17 Spent Fuel and Uranium Resources

18 Fast Neutron Reactors A solution for both an optimized use of resources and waste minimization Fast neutrons allow a more efficient burning of actinides because the ratio of fission/capture cross sections is higher than with thermal neutrons. The first consequence is the possibility of positive breeding gains which allows to burn all the uranium through conversion of Uranium 238 in Plutonium 239. Another interesting feature is the possibility of burning all the actinides produced in the fast reactors themselves or in light water reactors by continuous recycling, thereby reducing considerably the long term radioactive potential of waste. 18

19 Breeding Process If a neutron is captured by a Uranium-238, following a short series of decays, it is transformed to Plutonium-239. The process is shown in the figure below: symbol +n indicates a neutron absorption, ß - represents radioactive decay by beta emission with the half-life shown below the arrow. 238 U +n 239 U β 239 Np β 239 Pu 23.5 min 2.35 days 19

20 Brutprozesse Brutprozesse Links U-238 Pu-239 Rechts Th-232 U

21 Recycling to optimize the use of uranium resources Fast neutron reactors burn plutonium while converting U-238 into plutonium that is burnt in situ (regeneration breeding of fissile fuel) The existing depleted uranium that is stored today in France is worth 5000 years of current nuclear production. 21

22 Waste minimization 22

23 Technologies Fast Reactors To keep a fast neutron system it is necessary to avoid light elements in the core and in particular for the cooling system. The two main possibilities for cooling are liquid metals or gases. They are part of Gen IV selected systems. A broad worldwide effort have been devoted to sodium cooling technology including industrial prototypes (BN600 in Russia, Superphenix in France, Monju in Japan). The Russians have used lead cooling for naval reactors and some more studies have been made for possible use of lead or lead-bismuth cooling systems. The use of helium technology developed for HTGRs is also considered for fast reactors. After a negative attempt to use vapor cooling, supercritical water has been selected by the GIF as a possible candidate for fast reactor cooling. 23

24 Fast Reactors in Operation/under Construction Japan Joyo experimental reactor (140 MWth) suspended Monju, prototype (280 MWe) expected to be decommissioned Russia BOR -60 experimental reactor (12 MWe) in operation MBIR, - experimental reactor (55MWe) under construction BN-600 prototype (600MWe) in operation BN-800 demonstration (800 MWe) in operation BN-1200 commercial (1220 MWe) under construction BREST MWe construction started 24

25 Fast Reactors in Operation/under Construction China CEFR, experimental (20 MWe) in operation CDFR-1000 demonstration (1000MWe) under construction India FBTR, experimental (20 MWe) in operation PFBR, prototype (500 MWe) under commissioning 25

26 Liquid Metal Fast Breader Reactors Liquid metals are the preferred option due to their excellent heat transfer properties Possible choices of liquid metal coolant are mercury, lead, sodium and a sodium-potassium (NaK) mixture Sodium best choice: High density: 0,85 g/cm³ at 400 C High specific heat : 1.28 J C/g High boiling point: 883 C Melting point: 98 C Strongly activated: Na-24, 15 h half-life Intense gamma emitter: 1,4 MeV 26

27 Core and Blanket The LFMBR core is composed of two parts: core and blanket The fission process takes place in the core volume Extra neutrons diffusing out from the core are absorbed in a material ( depleted U-238) surrounding the core which is called the radial blanket. In the vertical direction escaping neutrons are absorbed in the vertical blanket. This material is directly incorporated into each fuel rod above and below the fuel region (depleted U-238) 27

28 Physics behind LFMBR The average number of neutrons produced in one fission process is around η= 2.5 for thermal fission and increases up to about η =3 at 100 kev neutron energy Therefore LFMBR has no light nuclei in core (no H,D,C,O,Be) One neutron is necessary to continue chain reaction, some neutrons are lost in reactor materials Extra neutrons can be captured by U-238 to be converted to Pu-239 Total number of fissile nuclei (Pu-239) in the reactor increase as the reactor operates 28

29 Example Assumption: 100 fissions produce 300 fast neutrons neutrons to continue chain reaction neutrons convert U-238 to Pu-239 in the core - 40 are lost by parasitic absorption in core - 60 leave the core for the blanket (leakage) + 50 convert U-238 to Pu-239 in the blanket - 10 are lost again by parasitic absorption 29

30 Physics 2 Breeding ratio (BR): Number of produced fissile nuclei (Pu-239) by absorption in U-238 to number of consumed nuclei in fuel BR in core: 0.8 BR in blanket: 1.25 Main Pu-239 production in blanket due to resonance absorption in U-238 (between 5 and 5000 ev) Conversion factor: C = production rate of fissile atoms consumption rate of fissile atoms If C > 1 breeding ratio 30

31 Sodium Cooled Fast Reactor Pool Design

32 Sodium Cooled Fast Reactor Reactor Parameters Outlet Temperature Pressure Rating Fuel Cladding Average Burnup Conversion Ratio Average Power Density Reference Value C -1 Atmospheres MWth Oxide or metal alloy Ferritic or ODS ferritic GWD/MTHM MWth/m³

33 Sodium Cooled Fast Reactor System Fast-spectrum reactor and closed fuel recycle system. Power from a few hundred MWe to MWe. Sodium core outlet temperatures are typically ºC. Either pool layout or compact loop layout is possible. Primary system operates at essentially atmospheric pressure, pressurized only to the extent needed to move fluid. Sodium reacts chemically with air and with water, design must limit the potential for such reactions and their consequences Secondary sodium system acts as a buffer between the radioactive sodium in the primary system and the steam or water If a sodium-water reaction occurs, it does not involve a radioactive release.

34 Sodium Technology Sodium is a very suitable coolant: liquid in a wide range of temperatures ( C) mono isotope (Na-23) thermodynamics parameters no corrosion (when purified) Large industrial experience : various industrial uses 40 years of technological studies for nuclear applications many prototypes Well-known drawbacks : chemical reactivity (sodium fires and sodium-water reactions) difficulties for handling and inspection 34

35 Safety Features long thermal response time a reasonable margin to coolant boiling, a primary system that operates near atmospheric pressure an intermediate sodium system between the radioactive sodium in the primary system and the power conversion system 35

36 Actinides Managment SFRs operate with high energy neutrons that are more effective at fissioning actinides. The main characteristics of the SFR for actinide management mission are: Consumption of transuranics in a closed fuel cycle, thus reducing the radiotoxicity and heat load which facilitates waste disposal and geologic isolation. Enhanced utilisation of uranium resources through efficient management of fissile materials and multi-recycle. Minor Actinides: Np, Am, Cm, Bk, Cf, Es, Fm responsible for the bulk of radioactivity in waste!! 36

37 Pool Type Vessel very simple design with only few pipes Disadvantages of the pool: vessel is large, must be fabricated on-site, difficult quality assurance In operation internal structures difficult to inspect as they operate under liquid sodium 37

38 Phenix Reactor - France 38

39 Pool Design vs.loop Design 39

40 Loop Type Vessel much smaller Can be built in a factory and transported to the site The pipework of the loop reactor may be longer and more complicated but it is easier to inspect 40

41 SFR Prototype Monju Japan 41

42 Monju Japan First criticality: 8/1994 Shut down 1996 due to Na leak, some short operation periods until 2010 Planned to be decommissioned Cooling System: Na cooled (loop-type) Thermal output: 714 MWth Electrical output: 280 MWe Fuel: PuO2 + UO2 Plutonium enrichment Inner core: 16 % Pu 239 Outer core: 21% Pu 239 Core Dimensions: Diam./Height: 180/93 cm Volume: 2340 litres 42

43 Reactor Core Breeder Core composed of three regions: Inner core: U-235+Pu-239 Outer core: U-235+Pu-239 (higher enrichment) Breeding blanket (U-238) Reflector elements not shown 43

44 View on Core Structure 44

45 Fuel Assembly Hexagonal: 169 rods per assembly, in centre 92 cm fuel zone (U-235+Pu239), above and below breeder zone (U-238 only) Fuel rod: 2.8 m long, 6.5 mm diameter Fuel Mass: Core ( U+Pu metal ): 5.9 t Blanket ( U metal ): 17.5 t Number of fuel assemblies: Inner core: 108 Outer core: 90 Cladding material: SS 316 Power density: 275 kw / litre 45

46 Breeder Blanket Radial breeder assemblies: 172 Blanket thickness: Upper: 30 cm Lower: 35 cm Radial: 30 cm Control rods: Fine: 3 Coarse: 10 Back-up: 6 Breeding ratio: 1.2 approx. 46

47 Cooling Circuits Primary sodium temperature: Reactor inlet / outlet: 397 C / 529 C Secondary sodium temperature: IHX inlet / outlet: 325 C / 505 C Number of loops: 3 Reactor vessel dimensions: Height / diameter: 18 / 7 m Steam Pressure: 127 kg/cm² Interval between refuelling: 6 months Refuelling system: Single rotating plug with fixed arm fuel handling machine 47

48 Superphenix 1200 MWe plant built at Creys Malville, in operation

49 Superphenix Criticality: 9/1985 Shut down: 12/ MWe net 3000 MWth Efficiency: 41.3% 49

50 Superphenix Heat Transfer System Primary (Na) Coolant Circuit: Total amount: 3250 tons Core inlet temperature: 395 C Core outlet temperature: 545 C Inlet intermediate heat exchanger (IHX): 525 C Secondary (Na) Coolant Circuit: Total amount: 1500 t IHX outlet: 525 C IHX inlet: 542 C Water Steam Circuit: Steam Generator inlet temperature: 525 C Steam Generator outlet temperature: 490 C 50

51 Heat Transfer System 51

52 Heat Transfer System Number of primary sodium pumps: 4 Number of intermediate (Na/Na) heat exchangers (IHX): 8 Number of secondary sodium pumps: 4 Number of steam generators: 4 Number of feedwater pumps: 4 52

53 Technical Data Fuel assemblies: Number of fuel assemblies: 364 Total length: 5.4 m Active length: 1.95 m Number of rods per assembly: 271 Outer diameter fuel rod: 8.5 mm Fuel: MOX 15%UO2, 85%PuO2 Maximum burn up: ca MWd/ton Cladding: Stainless steel 53

54 Technical Data Breeding assemblies: Number of fuel assemblies: 233 Total length: 5.4 m Active length: 1.95 m Number of rods per assembly: 91 Outer diameter fuel rod: 10.5 mm Material: Depleted U-238 Cladding: Stainless steel 54

55 Shut Down System Primary shut down system: Number of primary absorber elements: 21 Number of absorber rods per absorber element: 31 Absorber material: Borated steel Absorber length: 1.3 m Secondary shut down system: Number of secondary absorber elements: 3 Number of absorber segments per absorber element: 3 Absorber material: Boron carbide 55

56 Reactor Core 56

57 Fuel Assembly Overall length: 5.4 m Active length: 1.95 m Total number of assemblies in core:354 Number of rods per assembly: 271 Cladding material: SST Maximum cladding temperature: 620 C 57

58 Refuelling 58

59 Past and present LFMBRs 59

60 Past and Present LFMBRs 60

61 Russian Breeder Projects The BN-800 Beloyarsk-4 fast reactor will replace the BN-600 unit 3 at Beloyarsk, 790 MWe, started in mid 2014 It sometimes considered as first Generation III+ reactor which, after 2020, will start to take a large share of Russian capacity Fast reactors should supply some 14 GWe by 2030 and 34 GWe of capacity by The BN-1200 Beloyarsk-5 reactor is expected to be complete by 2025 at Beloyarsk 61

62 BN 800 BN 800 is a 3 loop pool type FBR. Flexible on fuel (U+Pu, MOX or metal) breeding ratio at 1 Will burn 40 tons of seperated Pu by 2025 No breeding blanket implemented 910 tons Na, outlet primary temperature is 547 C Closed fuel cycle actinides that are produced are consumed in reactor (Pu and minor Actinides) 62

63 BN 1200 Start up in loop design, 2800 MWth, 550 C, 60 years lifetime Sodium cooled but lead coolant is also possible Efficiency. 43,5% gross, 40,7 % net Breeding ratio: 1.2 with MOX fuel, 1.35 or 1.45 with nitrite fuel 426 fuel assemblies 174 radial blanket assemblies, surrounded by 599 boron shielding assemblies 63

64 Lead-Cooled Fast Reactor System

65 Lead Cooled Fast Reactor Reference values Reactor Parameters Reference Value Pb-Bi Battery (nearer-term) Pb-Bi Module (nearer-term) Pb-Large (nearer-term) Pb-Battery (nearer-term) Coolant Outlet Temperature( C) Pressure(Atmospheres) Power approx (MWth) Fuel Cladding Average Burnup (GWD/MTHM) Conversion Ratio Lattice Primary Flow Pb-Bi Metal Alloy or Nitride Ferritic Open Natural Pb-Bi Metal Alloy Ferritic d>1.0 Open Forced Pb Nitride Ferritic Mixed Forced Pb Nitrid Ceramic coatings or refractory alloys Open Natural

66 Lead-Cooled Fast Reactor System LFR systems are Pb or Pb-Bi alloy-cooled reactors with a fastneutron spectrum and closed fuel cycle. Options include a range of plant ratings, including a long refuelling interval battery ranging from MWe, a modular system from MWe, and a large monolithic plant at 1200 MWe. It had the highest evaluations to the Generation IV goals among the LFR options, but also the largest R&D needs and longest development time. The nearer-term options focus on electricity production and rely on more easily developed fuel, clad, and coolant combinations and their associated fuel recycle and re-fabrication technologies. The long term option seeks to further exploit the inherently safe properties of Pb and raise the coolant outlet temperature sufficiently high to enter markets for hydrogen and process heat, possibly as merchant plants.

67 BREST 1 Reactor Coolant Pump 2 CPS column on rotary plug 3 Steam water headers 4 Core barrel 5 Steam Generator 6 Core 7 Vessel Commissioned: 2020 Power: 300 MWe Fuel:UN and PuN Core inlet/outlet lead temp: 420/450 C SG inlet/outlet water /steam temp: 340/505 C Efficiency: 43%

68 Advantages As no electricity is required for the cooling after shutdown, this design has the potential to be safer than a water-cooled reactor. Liquid lead-bismuth systems can't cause an explosion and quickly solidify in case of a leak, further improving safety. Lead is very dense, and therefore a good shield against gamma rays. Lead's nuclear properties allow it to prevent a positive void coefficient, which is difficult to prevent in large sodium fast reactor cores. Does not neet backup power, natural convection for decay heat removal is enough 68

69 Disadvantages /Challenges Lead and lead-bismuth are very dense, increasing the weight of the system therefore requiring more structural support and seismic protection which increases building cost. Solidification of the lead-bismuth solution renders the reactor inoperable. However, lead-bismuth eutectic has a comparatively low melting temperature of C, making desolidification a relatively easily accomplished task. Lead has a higher melting point of С, but is often used as a pool type reactor where the large bulk of lead does not easily freeze. Pb can lead to Po production Opacity makes inspections difficult Corrosive at high temperatures when in contact with structural steels 69

70 Gas Cooled Fast Reactor

71 Gas Cooled Fast Reactor Reactor Parameters Reference Value Reactor Power Net plant efficiency (direct cycle helium) Coolant inlet/ outlet temperature and pressure Average power density Reference fuel compound Volume fraction, Fuel/ Gas/ SiC Conversion ratio 600 MWth 48% 490xC/850xC at 90 bar 100 MWth/m³ UPuC/SiC(70/30%)with about 20% Pu content 50/40/10% Self sufficient

72 Gas-Cooled Fast Reactor System The GFR system features a fast-spectrum helium-cooled reactor and closed fuel cycle. The high outlet temperature of the helium coolant makes it possible to deliver electricity, hydrogen, or process heat with high conversion efficiency. The GFR uses a direct-cycle helium turbine for electricity and can use process heat for thermo-chemical production of hydrogen. The GFR s fast spectrum also makes it possible to utilize available fissile and fertile materials (including depleted uranium from enrichment plants) two orders of magnitude more efficiently than thermal spectrum gas reactors with once-through fuel cycles. The GFR reference assumes an integrated, on-site spent fuel treatment and re-fabrication plant.

73 He Technology Gas cooling is less efficient than liquid metal cooling The development of a gas cooled fast reactor will require a new type of fuel Helium technology is already considered for VHTR Specific safety concerns which should be clarified If it can be successfully designed, the result will satisfy both objectives for a sustainable development (fast neutron physics and high temperature technology) 73

74 Supercritical Water Cooled Reactor

75 Supercritical Water Cooled Reactor Reactor Parameters Plant capital cost Unit power and neutron spectrum Net efficiency Coolant inlet and outlet temperature and pressure Average power density Reference fuel Fuel structural materials Cladding structural materials Burnup / Damage Safety approach Reference Value $ 900/KW 1700 MWe; thermal spectrum 44% 280xC/510xC/25Mpa 100MWth/m³ UO 2 with austenitic or ferritic martensitic stainless steel or Ni alloy cladding Advanced high strength Metal alloys are needed 45GWD/MTHM; Similar to ALWRs

76 Supercritical-Water-Cooled Reactor System SCWRs are high-temperature, high-pressure watercooled reactors that operate above the thermodynamic critical point of water (374 C, 22.1 MPa) These systems may have a thermal or fast-neutron spectrum, depending on the core design. SCWRs offer increases in thermal efficiency relative to current-generation LWRs. The efficiency of a SCWR can approach 44%, compared to 33 35% for LWRs. A lower-coolant mass flow rate per unit core thermal power. This offers a reduction in the size of the reactor coolant pumps, piping, and associated equipment, and a reduction in the pumping power.

77 Advantages High plant thermal efficiency (use of supercritical turbine generator system like in coal plants)-44% Simplification of primary system eliminiation of steam generators, pressurizers and main circulation pump (PWR), inner circulation pump, steam seperators and dryers(bwr) No phase change of water Zr can t be used as cladding no H production Enrichment 5 6 % - higher burnup possible Same passive safety features as Simplified BWR (cooling through natural convection) 77

78 Combination of PWR and BWR 78

79 Challenges Fuel cladding material Reactor structure material both need to withstand high temperatures Containment should withstand the high pressures that occur within reactor Peak reactivity during control rod insertation 79

80 Challenges for R&D Validation of transient heat transfer models (for describing the depressurization from supercritical to subcritical conditions), qualification of materials (namely advanced steels for cladding) demonstration of the passive safety systems. 80

81 High Temperature Gas Cooled Reactors (HTR)

82 Development of Helium Reactor Technology From 1963 to 1989 DRAGON (U.K.) AVR (FRG) PEACH BOTTOM 1 (U.S.A.) FORT ST. VRAIN (U.S.A.) THTR (FRG)

83 High Temperature Gas-cooled Reactors Potential to participate in the complete energy market with cogeneration and high temperature process heat application Significantly improved safety Position close to markets or heat users Can achieve higher fuel burnup ( GWd/t) Higher ( 20-50%) efficiency in electricity generation than conventional nuclear plants due to higher coolant outlet temperatures 83

84 Why HTGR? Non-Electric Applications Very high temperature reactors Gas-cooled fast reactors Molten Salt reactors Supercritical water-cooled reactors Sodium-cooled fast reactors Liquid metal cooled reactors Water cooled reactors District heating ( o C) Seawater desalination Pulp & paper manufacture Methanol production Heavy oil desulfurization Petroleum refining Methane reforming hydrogen production Thermochemical hydrogen production Coal gasification 84 Blast furnace steel making

85 Technical Background 1 For higher thermal efficiency higher gas temperatures up to 1000ºC are necessary CO 2 gas temperature is limited to 800ºC then it disintegrates into C and O Only stable gas with good nuclear and thermal properties is Helium No activation Chemically inert High specific heat Stable to high temperatures

86 Technical Background 2 For HTR no metallic fuel cladding possible Only solution is high density graphite Low neutron absorption Minimal radiation damage Superb heat resistance High thermal conductivity Therefore development of new fuel types - coated particles embedded in graphite Two major lines: 1. Graphite fuel spheres (pebbles): German version 2. Graphite prismatic fuel assemblies: US version

87 Advantages of HTR High efficiency: 47% Possibility of U-233 breeding from Th-232 Excellent passive safety features Strong negative temperature coefficient Excellent fission product retention Simple design Economically attractive Burn-up of MWd/ton

88 International Activities 1 Germany: 15 MW e AVR operated first HTR from 1967 to 1988, followed by 300 MW e Thorium-Hochtemperaturreaktor THTR (Pebble Bed Reactor) operated between 1986 to 1989 German development was stopped in the late 80-ties, but German industry cooperates with South Africa, China, Indonesia and Russia on modular units of Pebble Bed Modular Reactors (PBMR)

89 HTR cross cut

90 International Activities 2 South Africa plans 165 MW th PBMR between after 2012 China operates a 10 MW th (2 MW e prototype PBMR, critical since 2000 Japan works on a 40 MW th PBMR with prismatic fuel assemblies Russia plans a 330 MW e Pu Burner with prismatic fuel assemblies for 2014

91 High Temperature Pebble Bed Modular Reactor (PBMR) 360,000 pebbles in core About 3,000 pebbles handled by fuel handling system each day About 350 discharged daily One pebble discharged every 30 seconds Average pebble cycles through core 15 times Burn-up of MWd/t Fuel handling most maintenance-intensive part

92 Typical PBMR Technical Data 110 MW e Helium Cooled Indirect Cycle 8% Enriched Fuel Built in 2 Years Factory Built Site Assembled On-line Refueling Modules added to meet demand - no Reprocessing High Burn-up >90,000 MWd/t Direct Disposal of High Level Waste

93 PBMR Fuel Spheres Fuel particles (kernels) consist of uranium dioxide Coated particles embedded in a spherical graphite matrix, 50 mm diameter 5 mm thick fuel-free outer graphite zone, overall outer sphere diameter 60 mm One fuel sphere contains approximately coated particles fuel spheres are required for a single core loading

94 Composition of Graphite Spheres FUEL ELEMENT DESIGN FOR PBMR 5mm Graphite layer Coated particles imbedded in Graphite Matrix Dia. 60mm Fuel Sphere Pyrolytic Carbon 40/1000mm Silicon Carbite Barrier Coating Inner Pyrolytic Carbon 40/1000mm Porous Carbon Buffer 95/1000mm 35/1000mm Half Section Dia. 0,92mm Coated Particle Dia.0,5mm Uranium Dioxide Fuel

95 TRISO Fuel Particle Microsphere Pyrolytic carbon: Material similar to graphite with some covalent bonding between it layers Produced by heating CH 4 nearly to its breakdown temperature and permitting the graphite to crystallise on the UO 2 kernels

96 Ceramic Coated Fuel Particles

97 Microcut through Fuel Kernel

98 Production of Coated Particles

99 Fuel Sphere Production 1 Solution containing 8% enriched uranium is cast to form microspheres. These are washed, calcinated and sintered at high temperature to produce uranium dioxide kernels Kernels are then put in a Chemical Vapour Deposition furnace at a temperature above 1000ºC in which the pyrolytic carbon coating layers are added with extreme precision by cracking CH 4 into C+H SiC-layers have different densities: -one inner layer low density to capture volatile fission products, -two outer layers with high density

100 Fuel Sphere Production 2 Carbon densities can be adjusted by cracking temperature and exposure time The coated particles with about a millimeter in diameter are mixed with a resin and graphite powder and pressed into 50mm diameter spheres A 5mm thick layer of fuel free graphite powder is then added to form the "fuel-free" outer zone. The resulting spheres are then machined, carbonized and annealed

101 TRISO Fuel Particle (Kernels) Microsphere 0.9 mm diameter ~11,000 particles in every pebble Fission products retained inside microsphere TRISO acts as a pressure vessel (= first and second barrier) Reliability Defective coatings during manufacture ~1 defect in particles

102 Cross Section of the German 300 MWeTHTR Core operating from 1983 to 1988

103 THTR Core during loading procedure

104 US Fort Saint Vrain HTR

105 US Prismatic Fuel

106 China s HTR Programme In February 2006, the State Council announced that the small high-temperature gas-cooled reactor (HTR) was the second of two high priority projects for the next 15 years. The small HTR-PM units with pebble bed fuel were to be 200 MWe reactors, similar to that being developed in South Africa Plans have evolved to make them twin 105 MWe units driving a single steam turbine.

107 Reactor Building of China s HTR-10

108 HTR-10 ATW 12/ decision to construct HTR construction started after safety review December 2000 first criticality January 2003 full power operation Excellent operation record Used for reactor safety experiments during normal and abnormal operation

109 Operational Data of the HTR-10 ATW 12/2006 Design Operation Nuclear power [MW] Electric power [MW] Helium inlet temperature [ 0 C] Helium outlet temperature [ 0 C] Helium pressure [MPa] 3 3 Feed water temperature [ 0 C] Steam temperature [ 0 C] Feed water flow [kg/s] ,57 Steam pressure [MPa] Fuel number [-] 13,622 Graphite spheres number [-] 16,387

110 Chinese Modular HTR Modular Pebble Bed HTR Power 10 MWth Core : pebbles diameter 1,8m approximative height 2m Construction 1995 Criticality 1/12/ / Passive Decay Heat Removal

111 Core of HTR-10

112 Passive Safety Feature in Case of LOCA Fuel temperatures remain below design limits even during loss of cooling

113 Temp ( C) Loss of Coolant Flow Temperature well below design limits 265 MW PBMR Ref. Core: Temperature Distribution during a DLOFC 1600 Maximum Fuel Temperature Average Fuel Temperature Maximum RPV Temperature Average RPV Temperature Time (h)

114 Chinese Pebble Bed-HTR PM 200 MWe demonstration unit at Weihei/Shandong province Later identical modules are planned Project costs US 375 M$ Start-up in fuel elements, 9% enriched Low power density but high temperature 900 C, efficiency 40% 60 years life time, 85% availability First of 18 modules

115 Parameters of the HTR_Pebble Bed Module (HTR-PM) ATW 12/2006 Reactor thermal power [MW] Electric power [MW] 458 MW 195 MW Fuel number at equilibrium core [-] 520,00 Core Heigh/out diameter/inner diameter [m] 11/ 4 / 2.2 Helium pressure [MPa] 9 Helium flow rate [kg/s] 176 Helium temperature inlet/outlet [ 0 C] 250/750 Steam flow rate [t/h] Steam temperature [ 0 C] 538 Steam temperature [MPa] 13.5

116 RPV

117 Future Programs on HTR Technology in China Commercialization Duplication, mass production Next step project: Super critical steam turbine, co-generation R&D on future Technologies Higher temperature Hydrogen production Process heat application Gas turbine

118 Mock-up of Chinese HTR-Plant (Blocks of 105 MWe each) in Shangdon Modules based on the HTR-10 experience may be added according to needs (Jan 2016)

119 What you should remember Only Helium as coolant for HTR possible Helium has excellent cooling properties High thermal effciency Graphite as moderator and fuel cladding in form of coated particles Spherical or prismatic fuel elements Originally developed in USA and Germany Now followed by China and Russia

120 VHTR Thermal spectrum, once-through uranium cycle Prismatic block or pebble bed fuels Highly ranked in economics & in safety and reliability Hydrogen production & other process-heat applications 120

121 VHTR 121

122 Iodine Sulfur Process for H prodcution 122

123 Very High Temperature Reactor Reactor Parameters Reactor power Coolant inlet/outlet temperature Core inlet /outlet pressure Helium mass flow rate Average power density Reference fuel compound Net plant efficiency Reference Value 600 MWth 640 / 1000 C Dependent on process 320 kg/s 6 10 MWth/m³ ZrC coated particles in blocks, pins or pebbles > 50 %

124 Very High Temperature Reactor Systems Graphite-moderated, helium-cooled reactor with thermal neutron spectrum able to supply nuclear heat, core-outlet temperatures of 1000 C, either prismatic block core or a pebble-bed core. VHTR may produce hydrogen from only heat and water using thermo-chemical iodine-sulfur (I-S) process Possible daily production over 2 million normal cubic meters of hydrogen. It may also generate electricity with high efficiency, over 50% at 1000 C, compared with 47% at 850 C in the GTMHR or PBMR. High core outlet temperatures would enable nuclear heat application to such processes as steel, aluminium oxide, and aluminium production. Direct cycle electricity generation with the helium gas turbine is possible. For nuclear heat applications such as process heat for refineries, petro-chemistry, metallurgy, and hydrogen production, the heat application process is generally coupled with the reactor through an intermediate heat exchanger (IHX), which is called an indirect cycle.

125 HTR Idaho 125

126 Molten Salt Reactor

127 Molten Salt Reactor Reactor Parameters Net power Power density Net thermal efficiency Fuel - salt - inlet temperature - outlet temperature - vapour pressure Moderator Power Cycle Neutron spectrum turner Reference Value 1000 Mwe 22 MWth/m³ 44 to 50% 565 C 700 C (850 C for H-Production) <0.1 psi Graphite Multi - reheat recuperative HE Thermal - actinide

128 Molten Salt Reactor System The MSR produces fission power in a circulating molten salt fuel mixture fuelled with uranium or plutonium fluorides dissolved in a mixture of molten fluorides, with Na and Zr fluorides as the primary option. MSRs have good neutron economy, opening alternatives for actinide burning and/or high conversion High-temperature operation holds the potential for thermo-chemical hydrogen production Molten fluoride salts have a very low vapour pressure, reducing stresses on the vessel and piping Inherent safety by fail-safe drainage, passive cooling low inventory of volatile fission products in the fuel Refuelling, processing, and fission product removal performed online, potentially yielding high availability MSRs allow the addition of actinide feeds to the homogenous salt solution without blending and fabrication needed by solid fuel reactors.

129 MSR Concept Epithermal or fast neutron spectrum possible Variety of fuels Th-232 is considered ( can breed U-233) Salts are consiedered as primary coolant (Li-Be Fluoride or Li Fluoride) C without pressurization 129

130 Function Fuel is a molten mixture of FLiBe with dissolved LEU (UF 4 ) Core consists of unclad graphite moderator salt can flow at 700 C Single fluid MSR or homogenous design: Th is dissovled with U Heterogenous or two fluid design: Fissile U or Pu in primary loop fertile salt contains Th in secondary loop Can operate as breeding reactor (MSBR) Reactor vessel needs to be removed after 8 years 130

131 Function Liquid MSR Fission products dissolve in fuel salt and are continously removed in reprocessing loop and replaced with fissile or fertile material; Cs and I nuclides remain in salt much higher burnup possible Less decay heat Freeze Plugs if exessive temperatures are reached the primary salt will be drained by gravity away from moderator into dump tanks Large negative void coefficient and temperature coefficient shut down automatically at higher Power levels 131

132 Historical Development Some early experience: ORNL aircraft power plant 1953 Aircraft reactor experiment (ARE) 1954 Aircraft reactor test (ART) 1956 Molten Salt Reactor Experiment ( ) 8 MWt Single region core Graphite moderated Also used U-233 and mixed U/Pu salt fuel On-line refueling >13,000 full power hours 132

133 Current Development 2 fast spectrum MSR concepts are being studied, large power units based on homogeneous core with liquid fluoride-salt circulating fuel: MSFR design in France, Euratom and Switzerland MOSART concept in the Russian Federation. The US is working on solid fuel FHR (Fluoride-salt-cooled High-temperature Reactor) as well as liquid fueled MCFR (Molten Chloride salt Fast Reactor) China, is working on FHR and TMSR (Thorium Molten fluoride Salt thermal Reactor) designs 133

134 Current R&D efforts salt physical, chemical and thermodynamic properties, development of advanced materials, including studies on its compatibility with molten salts and behaviour under high neutron fluxes at high temperature, mastering of corrosion and tritium release prevention technologies, basing on proper salt Redox control, development of efficient techniques of gaseous fission products extraction from the fuel salt by He bubbling, fuel salt processing flowsheet, including reductive extraction tests (actinide-lanthanide separation), system design, including development of advanced neutronic and thermal hydraulic coupling models, development of a safety, safeguards, security and proliferation resistance approaches dedicated to liquid fuelled reactors. 134

135 Challenges Higher containment radiation environment Safety requirements and licensing framework does not exist Limited experience for commercial demonstration Graphite material challenges (in coolant / fuel salt) and lifetime issues (high fluences) 135

136 Safety, Safeguards, and Security Context for Molten Salt Reactors MSR safeguards and safety will present new challenges for the designer and regulator Safeguards approaches may require a major shift from current practices Designer/Safeguard experts need to get involved early in order address the challenges MSR safety approach requires early interaction with the regulator May require new policies and certainly a modification to regulations/requirements - risk informed performance based approaches Use of simulation and modeling, separate effects testing, test and demonstration/prototype reactors 136 Status of MSR development

137 Proliferation Resistance Has Become Dominant Concern For All Fuel Cycles MSRs can be highly proliferation resistant or vulnerable depending on the plant design MSR designs until the mid-1970s did not consider proliferation issues Several current MSR design variants do not include separation of actinide materials Liquid fuel changes the barriers to materials diversion Lack of discrete fuel elements combined transmutation prevents simple accounting with continuous Solid LEU fresh fuel salt in transport and storage accountancy resembles LWR fuel Homogenized fuel results in an undesirable isotopic ratio a few months following initial startup (no short cycling) Extreme radiation environment near fuel makes changes to plant configuration necessary for fuel diversion very difficult High salt melting temperature makes ad hoc salt removal technically difficult Low excess reactivity prevents covert fuel diversion 137

138 MSFR CNRS France Based on this new design of the MSFR system, a preliminary accident list of the fuel circuit and the emergency draining system has been identified thanks to: The analysis of the accident type identified for current operated reactors (PWR); Accidental transient calculations of the MSFR with dedicated tools; Preliminary systemic risk analyses with a qualitative reevaluation to take into account new design. 138

139 Severe Accident Management Rupture of the Main Fuel Salt Pipe and Fuel Discharged on the Reactor Cell Bottom : The model based on mass transfer theory describing main radionuclides distribution between the fuel salt, metallic surfaces of the primary circuit, graphite and the gas purging system was applied for calculation releases to the containment atmosphere of MOSART For MSR the total release of radioactivity would be significantly lower (by I - 2 orders of magnitude compared to PWR), though for several particular nuclides such I131 and I133 the differences are smaller 139

140 Th R&D in China TMSR team has been trying to design and construct a saltcooled solid fuel test MSR (TMSR-SF1) and a Th-fueled test MSR (TMSR-LF1), plus a simulator of the molten salt-cooled reactor (TMSR-SF0). By the end of 2016 TMSR team: (1) collected information of the Haiyang site and evaluated the environmental impact of building the test MSR at Haiyang; (2) completed preliminary engineering design of the instrument and control and cover gas systems of TMSR-SF1; (3) prepared a draft PSA report of TMSR-SF1 based on the current preliminary engineering design and prepared for the application of the construction permit; (4) released conceptual designs of 2 MWt Th-fueled test TMSR-LF1 and small modular thorium-fueled demonstration TMSR-LF2. 140

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