Molten Salt Reactors

Transcription

1 Nuclear Engineering Panel Technical Presentation Molten Salt Reactors Date: Wednesday, 19 th March 2014 Time: 5.30 pm for 6.00 pm Venue: Engineers Australia Harricks Auditorium, Ground Floor, 8 Thomas St, Chatswood Speaker: Mark Ho, BEng, MEng ANSTO ABSTRACT: Molten salt reactors (MSRs) hold the promise of improved reactor safety, fuel utilisation and economy. The coolant, a chemically non-reactive fluoride-based molten salt (lithium -berylliumfluoride or FLiBe ), has a boiling temperature of 1400 C, far in excess of normal operational temperatures of about 650 C, giving the design a high degree of inherent safety. Also, the salt solidifies at 459 C, meaning any reactor shutdown would immobilize both fuel and fission products within the salt. The United States of America pioneered molten salt technology in the 1960s and built a liquid-fuelled 8 MW (thermal) test reactor in Oak Ridge National Laboratories (ORNL). The experiment demonstrated molten salt reactors inherent safety and ease of operation. After the molten salt reactor program was shut down in 1973, with preference being given to sodium fast breeder reactors, there has been little activity until now. ORNL has revived molten salt technologies in the past decade, detailing their work in a series of reports. More recently, the Shanghai Institute of Nuclear Applied Physics (SINAP), working in collaboration with ORNL, has a US$350 million, 5- year programme to build two molten salt reactor prototypes with solid and liquid fuels. This talk will review past and present developments in MSRs, highlighting the design s function and safety. PERSONAL DETAILS: Mark Ho has been working at ANSTO for 8 years as a numerical analyst specialising in reactor heat-transfer. He holds a BEng (Mech), and a MEng (Mech) for research in reactor fuel assembly flow-induced vibration phenomena. He is looking to submit his PhD (Mech, part-time) this year on computational fluid dynamics. His PhD work is on the modelling of deformable 3D volumes and surfaces for the simulation of bubbly flows. He also has a strong interest in advanced power reactor designs, in particular molten salt reactors for their potential to supply nuclear power for low-carbon electricity generation, even more safely than current nuclear plants.

2 A brief introduction on Molten Salt Reactors Mark Ho, Nuclear Analysis Section, Centre for Nuclear Applications, ANSTO. Molten Salt Reactor Experiment, MW thermal U-233, U-235 fuel dissolved in salt Adv. High Temp. Reactor, (AHTR concept) 3,400 MW thermal 1,500 MW electric Salt cooled with solid TRISO fuel

3 Molten Salt Reactors 1. Protons, neutrons, electrons. 2. Quick comparison between a PWR and MSR. 3. ANP (Aircraft Nuc. Propulsion Program) 4. MSRE (Molten Salt Reactor Experiment) 5. SINAP (Shanghai Inst. Nuc. Applied Physics) 6. AHTR (Advanced High Temperature Reactor)

4 The atomic structure (simplified). Source: APRANSA

5 Pressurised Water Reactor - about 420 units worldwide including BWRs. 1. Fuel: uranium dioxide (4-5% enriched) 2. Moderator: light water 3. Coolant: light water 4. Materials Vessel / piping: stainless steel, Inconel. Control rod: cadmium, silver, indium. Cladding: zircaloy (98% zirconium, neutron transparent) 5. Neutron shielding: light water, concrete. 6. Neutron reflector: light water. Graphics: Nuclear Regulatory Commission. (NRC) Operating temperature: C Operating pressure: 150 atm Rankine (steam) cycle: 33 % efficiency.

6 Molten Salt Reactor Experiment - unique 8 MW (thermal) design, ORNL, Graphics: Oak Ridge National Labs. (ORNL) 1. Fuel: uranium tetrafluoride (dissolved ) ( 235 U, 35% enriched, 233 U added later) 2. Moderator: graphite 3. Coolant: lithium-beryllium-fluoride (FLiBe). 4. Materials Vessel / piping: nickel alloy, graphite Control rod: Gd 2 O 3 - Al 2 O 3 Cladding: none, (fuel in solution with salt!). 5. Neutron shielding: concrete. 6. Neutron reflector: graphite Operating temperature: C (!) Operating pressure: 1 atm (!) No turbine installed but high temp. can drive: Brayton cycle: 45 % efficiency or Supercritical water cycle: 40% efficiency

7 Aircraft Nuclear Propulsion Programme. HTRE- 1 and 2 HTRE- 3 Fuel = 235 Uranium, high enrichment. Moderator = Beryllium Oxide (BeO) Coolant = NaF - ZrF 4 - UF 4 ( mol %) Power = 35 MW (HTRE - 3) Operating hours = 126 hours ( HTRE - 3 in ) Main research conducted at Oak Ridge National Laboratory (ORNL) Payroll: 600 on the Aircraft Reactor Experiment in supporting research. We didn t think it would work but we didn t turn down the funding ~ Weinberg

8 The Molten Salt Reactor Experiment ( ) Reactor systems: 1. Reactor Vessel 2. Heat Exchanger 3. Fuel Pump 4. Freeze Flange 5. Thermal Shield 6. Coolant Pump 7. Radiator 8. Coolant Drain Tank 9. Fans 10. Fuel Drain Tanks 11. Flush Tank 12. Containment Vessel 13. Freeze Valve

9 Molten Salt Reactor Experiment ( ) Fuel Salt Pump Motor Nuclear Reactors using FLiBe were successfully built and operated MSRE Reactor Vessel Heat Exchanger Molten-Salt Reactor Experiment ( )

10 How does molten salts compare with water? H 2 O N/A Source: ORNL Report: TM Assessment of Candidate Molten Salt Coolants for the AHTR

11 FLiBe salt. Melts at 459 C boils at 1400 C Dissolves 232 Th, 238 U, 233 U, 235 U or 239 Pu 233 U can be dissolved into the salt coolant to minimise handling. Advantages: Ease of containment, excellent heat capacity. Neutron economy and moderation. Potential for high fuel burn-up >50%, not 5%. Negative thermal reactivity coefficient. Xenon out-gassing possible. chemically inert. Some challenges: enrichment of 7 Li required to minimise tritium production by 6 Li. Beryllium is chemically toxic.

12 The molten salt breeder reactor (MSBR) started 1970, cancelled Dr. Alvin Weinberg, co-inventor of PWRs and MSRs, Director of ORNL The MSRE ran successfully for 9000 hrs on both 233 U and 235 U. The next stage was to construct a MSBRs for breeding U-233 from thorium. However, studies by ORNL showed the breeding ratio of 1.03 was low compare to the breeding ratio of 1.3 for sodium fast reactors under study by Idaho Nat. Lab. The Nixon administration, under the advise of the AEC, decided to divert all funding to the EBR-2. Little work has been done on in MSRs until now.

13 Revival of Molten Salt Reactors Prof. Charles Forsberg, MIT Director, P/I of AHTR Project Dr. Jiang Mianheng Head of CAS, Shanghai Branch Dr David Le Blanc Terrestial Energy (Canada) Dr. Cecil Park, ORNL Director of Reactor and Nuc. Systems Dr. Xu Hong Jie, Director of SINAP Mr Kirk Sorensen, FLiBe Energy (USA)

14 Courtesy of Prof. C. Forsberg, MIT. SINAP s work.

15 Chinese Academy of Science s (CAS) Plan. Courtesy of Prof. C. Forsberg, MIT.

16 Two FHR paths forward China s Strategy impacts US strategy. Courtesy of Prof. C. Forsberg, MIT.

17 Advanced High Temperature Reactor AHTR (also known as FHRs)

18 Integrating the best technologies into the AHTR Gas Cooled Reactors TRISO fuel Structural ceramics High temperature power conversion Molten Salt Reactors Fluoride salt coolant Structural alloy Hydraulic components Liquid Metal Reactors Passive decay heat removal Low pressure design Hot refueling Fluoride Salt Cooled Reactors High temperature Low pressure Passive safety Light Water Reactors High heat capacity coolant Transparent coolant Advanced Coal Plants Supercritical water power cycle Structural alloys The AHTR program is funded at $7.5 million / yr Activities divided between ORNL, UC Berkeley, MIT, U Wisconsin M. and Westinghouse. Graphics: ORNL.

19 American outlook on FHRs FHRs are characterised by: (1) thermal efficiency, (2) low-pressure operation, and (3) passive safety. Use of fluoride salts prevents major offsite radionuclide releases, even in the case of beyond-design-basis accident. Plans to use Nuclear Air-Brayton-Combined-Cycle (NACC) Technology to deliver hybrid nuclear-base-load and gas-turbine-variable-load electricity generation. The implications are: Environmental: It will allow a low-carbon-nuclear / renewable electricity grid Economic: Increase revenue (projected at 50%) relative to base-load nuclear power. Risk minimised technological development path, using as much commerical-off the shelf technologies. The purpose of the AHTR program is not to reproduce the MSRE, it s to test the technologies required for licensing a 60 yr molten salt power reactor.

20 Fluoride High-temperature Reactor (FHR) concepts are being developed for diverse applications AHTR (1500 MWe) PB-FHR (410 MWe) MIT currently designing MSR pilot-plant reactor for systems integration testing. Including: (1) materials irradiation performance (2) tritium control (3) reliable operations etc. SmAHTR (125 MWt) AHTR = Advanced High Temperature Reactor PB-AHTR = Pebble Bed Advanced High Temperature Reactor SmAHTR = Small Modular Advanced High Temperature Reactor Graphics courtesy of ORNL.

21 Tri-structural isotropic (TRISO) High Temperature Fuel Technology Fuel Particle AGR testing spans power density anticipated for FHRs Source: INL.

22 Coated Particle Plate Fuel Assemblies Coated particle fuel is a uranium oxy-carbide variant currently being qualified under DOE-NE Advanced Gas Reactor (AGR) program Fuel particles are configured into stripes just below the surface of the fuel plates Minimizes heat conduction distance to coolant Fuel plates have a 5.5 m fueled length Fuel assemblies are surrounded by a C-C composite shroud to channelize coolant flow Fuel Plate Cross Section Source: ORNL.

23 Primary to Intermediate Heat Exchanger Intermediate to Power Cycle Heat Exchanger Condenser Turbine Generator AHTR is Progressing Towards a Pre-conceptual Design Level of Maturity Both reactor and power plant systems are included in the modeling Decay Heat Cooling Tower Natural Draft Heat Exchanger Thermal Power Electrical Power Top Plenum Temperature Coolant Return Temperature AHTR Properties Number of loops MW 1500 MW 700 C 650 C Primary Coolant 7 LiF-BeF 2 Direct Reactor Auxilary Cooling System Heat Exchanger Core Vessel Pump Cooling Tower Fuel Uranium Enrichment Fuel Form Refueling UCO TRISO 9% Plate Assemblies 2 batch 6 month

24 PB-AHTR

25 SmAHTR is A Cartridge Core, Integral-Primary- System FHR Overall System Parameters Parameter Value Power (MWt) 125 Primary Coolant 7 LiF-BeF 2 Primary Pressure (atm) ~1 Core Inlet Temperature (ºC) 650 Core Outlet Temperature (ºC) 700 Core coolant flow rate (kg/s) 1020 Operational Heat Removal Passive Decay Heat Removal Reactor Vessel Penetrations 3 50% loops % loops None

26 FHR Safety Derives from Inherent Material Properties and Sound Design Inherent Large temperature margin to fuel failure Good natural circulation cooling Large negative temperature reactivity feedback High radionuclide solubility in salt Low pressure Source: ORNL Physor workshop Engineered High quality fuel fabrication Effective decay heat sinking to environment Passive, thermally driven negative reactivity insertion Multi-layer containment

27 Remaining Challenges for FHRs FHRs will use as many commercial-off-the-shelf technologies to minimise developmental risks and delays. Realising FHRs will require: Reactor systems testing and integration testing, ultimately in the form of a pilot-plant. 7 Lithium enrichment must be reindustrialized. Tritium extraction technology must be developed and demonstrated. Structural ceramics must become safety grade engineering material. Safety and licensing approach must be developed and demonstrated. Structured coated particle fuel must be qualified.

28 Size Comparison

29 Thanks for listening. Please ask me some questions.