A brief introduction on Molten Salt Reactors
Transcript of A brief introduction on Molten Salt Reactors
A brief introduction on
Molten Salt Reactors
Mark Ho, Nuclear Analysis Section,
Centre for Nuclear Applications, ANSTO.
Molten Salt Reactor Experiment, 1965
8 MW thermal
U-233, U-235 fuel dissolved in salt
Adv. High Temp. Reactor, (AHTR concept)
3,400 MW thermal 1,500 MWelectric
Salt cooled with solid TRISO fuel
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)
The atomic structure (simplified).
Source: APRANSA
http://www.arpansa.gov.au/radiationprotection/basics/atomic.cfm
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.
Operating temperature: 295 – 328 °C
Operating pressure: 150 atm
Rankine (steam) cycle: 33 % efficiency.
Graphics: Nuclear Regulatory Commission. (NRC)
Molten Salt Reactor Experiment - unique 8 MW(thermal) design, ORNL, 1965.
1. Fuel: uranium tetrafluoride (dissolved )
(235U, 35% enriched, 233U added later)
2. Moderator: graphite
3. Coolant: lithium-beryllium-fluoride (FLiBe).
4. Materials
Vessel / piping: nickel alloy, graphite
Control rod: Gd2O3 - Al2O3
Cladding: none, (fuel in solution with salt!).
5. Neutron shielding: concrete.
6. Neutron reflector: graphite
Operating temperature: 600 – 610 °C (!)
Operating pressure: 1 atm (!)
No turbine installed but high temp. can drive:
Brayton cycle: 45 % efficiency or
Supercritical water cycle: 40% efficiency
Graphics: Oak Ridge National Labs. (ORNL)
1947 - 1961
Aircraft Nuclear Propulsion Programme.
Fuel = 235Uranium, high enrichment.
Moderator = Beryllium Oxide (BeO)
Coolant = NaF - ZrF4 - UF4 ( 53 – 41 - 6 mol %)
Power = 35 MW (HTRE - 3)
Operating hours = 126 hours ( HTRE - 3 in 1958-1960 )
Main research conducted at Oak Ridge National Laboratory (ORNL)
Payroll: 600 on the Aircraft Reactor Experiment + 900 in supporting research.
“ We didn’t think it would work but we didn’t turn down the funding ” ~ Weinberg
HTRE- 1 and 2 HTRE- 3
The Molten Salt Reactor Experiment (1965-1969)
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
Nuclear Reactors using “FLiBe” were
successfully built and operated
MSRE
Reactor
Vessel
Fuel Salt
Pump Motor
Heat Exchanger
Molten-Salt Reactor Experiment (1965-1969)
Molten Salt Reactor Experiment (1965-1969)
How does molten salts compare with water ?
Source: ORNL Report: TM-2006 -12
Assessment of Candidate Molten Salt Coolants for the AHTR
H2O 18.0 0.0 N/A 1.0 1.00 1.0 0.58 75 246
FLiBe salt. •Melts at 459 °C
•boils at 1400 °C
•Dissolves 232Th, 238U, 233U, 235U or 239Pu
•233U 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 7Li required to minimise tritium
production by 6Li.
•Beryllium is chemically toxic.
The molten salt breeder reactor (MSBR)
started 1970, cancelled 1973.
• The MSRE ran successfully for 9000 hrs on both 233U
and 235U.
• 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….
Dr. Alvin Weinberg,
co-inventor of PWRs and MSRs,
Director of ORNL 1955 - 1973..
Revival of Molten Salt Reactors
Mr Kirk Sorensen,
FLiBe Energy (USA)
Prof. Charles Forsberg, MIT
Director, P/I of AHTR Project
Dr. Xu Hong Jie,
Director of SINAP
Dr David Le Blanc
Terrestial Energy (Canada) Dr. Jiang Mianheng
Head of CAS, Shanghai Branch
Dr. Cecil Park, ORNL
Director of Reactor and Nuc. Systems
SINAP’s work.
Courtesy of Prof. C. Forsberg, MIT.
Chinese Academy of Science’s (CAS) Plan.
Courtesy of Prof. C. Forsberg, MIT.
Two FHR paths forward
Courtesy of Prof. C. Forsberg, MIT.
China’s Strategy impacts US strategy.
Advanced High Temperature Reactor
AHTR
(also known as FHRs)
Integrating the best technologies into the AHTR
Fluoride Salt Cooled Reactors
• High temperature
• Low pressure
• Passive safety
Advanced Coal Plants• Supercritical water
power cycle• Structural alloys
Gas Cooled Reactors• TRISO fuel• Structural ceramics• High temperature power
conversion
Molten Salt Reactors• Fluoride salt coolant• Structural alloy• Hydraulic components
Light Water Reactors• High heat capacity
coolant• Transparent coolant
Liquid Metal Reactors• Passive decay heat
removal• Low pressure design• Hot refueling
Graphics: ORNL.
The AHTR program is funded at $7.5 million / yr
Activities divided between ORNL, UC Berkeley, MIT, U Wisconsin M. and Westinghouse.
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.
Fluoride High-temperature Reactor (FHR) concepts
are being developed for diverse applications
AHTR = Advanced High Temperature Reactor
PB-AHTR = Pebble Bed Advanced High Temperature Reactor
SmAHTR = Small Modular Advanced High Temperature Reactor
PB-FHR (410 MWe)
SmAHTR (125 MWt)
AHTR (1500 MWe) MIT currently designing MSR pilot-plant reactor
for systems integration testing.
Including: (1) materials irradiation performance
(2) tritium control
(3) reliable operations etc.
Graphics courtesy of ORNL.
Tri-structural isotropic (TRISO)
High Temperature Fuel Technology
Fuel Particle
AGR testing spans power density anticipated for FHRs
Source: INL.
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.
AHTR is Progressing Towards a Pre-conceptual
Design Level of Maturity
Both reactor and power plant systems
are included in the modeling
AHTR Properties
Thermal Power 3400 MW
Electrical Power 1500 MW
Top Plenum
Temperature
700 °C
Coolant Return
Temperature
650 °C
Number of loops 3
Primary Coolant 7LiF-BeF2
Fuel UCO TRISO
Uranium
Enrichment
9%
Fuel Form Plate Assemblies
Refueling 2 batch
6 month
Vessel
Core
Pump
Prim
ary
to
Inte
rmedia
te
Hea
t E
xch
ang
er
CoolingTower
Inte
rmed
iate
to
Pow
er
Cycle
H
ea
t E
xcha
ng
er
Ge
nera
tor
Turb
ine
Decay Heat Cooling Tower
Natural DraftHeat Exchanger
Direct ReactorAuxilary Cooling
System HeatExchanger
Con
den
ser
PB-AHTR
SmAHTR is A Cartridge Core, Integral-Primary-
System FHR
Parameter Value
Power (MWt) 125
Primary Coolant 7LiF-BeF2
Primary Pressure (atm) ~1
Core Inlet Temperature (ºC) 650
Core Outlet Temperature (ºC) 700
Core coolant flow rate (kg/s) 1020
Operational Heat Removal 3 – 50% loops
Passive Decay Heat Removal 3 – 0.25% loops
Reactor Vessel Penetrations None
Overall System Parameters
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
Engineered
• High quality fuel
fabrication
• Effective decay heat
sinking to environment
• Passive, thermally driven
negative reactivity
insertion
• Multi-layer containment
Source: ORNL Physor workshop 2012.
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.
• 7Lithium 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.
Size Comparison
Thanks for listening.
Please ask me some questions.