Nuclear Engineering Panel Technical Presentation

Molten Salt ReactorsDate: Wednesday, 19th March 2014Time: 5.30 pm for 6.00 pmVenue: Engineers Australia Harricks Auditorium,

Ground Floor, 8 Thomas St, Chatswood

Speaker: Mark Ho, BEng, MEngANSTO

ABSTRACT:

Molten salt reactors (MSRs) hold the promise of improved reactor safety, fuel utilisation andeconomy. The coolant, a chemically non-reactive fluoride-based molten salt (lithium-beryllium-fluoride or ‘FLiBe’), has a boiling temperature of 1400°C, far in excess of normal operationaltemperatures of about 650°C, giving the design a high degree of inherent safety. Also, the saltsolidifies at 459°C, meaning any reactor shutdown would immobilize both fuel and fission productswithin the salt. The United States of America pioneered molten salt technology in the 1960s andbuilt a liquid-fuelled 8 MW (thermal) test reactor in Oak Ridge National Laboratories (ORNL). Theexperiment demonstrated molten salt reactors’ inherent safety and ease of operation. After themolten salt reactor program was shut down in 1973, with preference being given to sodium fastbreeder reactors, there has been little activity until now. ORNL has revived molten salt technologiesin the past decade, detailing their work in a series of reports. More recently, the Shanghai Instituteof 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 willreview 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 reactorheat-transfer. He holds a BEng (Mech), and a MEng (Mech) for research in reactor fuel assemblyflow-induced vibration phenomena. He is looking to submit his PhD (Mech, part-time) this year oncomputational fluid dynamics. His PhD work is on the modelling of deformable 3D volumes andsurfaces for the simulation of bubbly flows. He also has a strong interest in advanced power reactordesigns, in particular molten salt reactors for their potential to supply nuclear power for low-carbonelectricity generation, even more safely than current nuclear plants.

A brief introduction on

Molten Salt Reactors

Mark Ho, Nuclear Analysis Section,

Centre for Nuclear Applications, ANSTO.

Molten Salt Reactor Experiment, 1965

8 MWthermal

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.