Overview of LFTR: Materials and Components

Kurt Harris, PhD Flibe Energy, Inc. kurt.harris@flibe-energy.com Nuclear Regulatory Commission Rockville, Maryland December 9, 2019 Three Nuclear Options The Thorium Reactor Company Located in Huntsville, Alabama Gen-4 Molten Salt Reactor Concept The MSRE successfully operated for over 20,000 hours, from 1965-1969. MSR Material Compatibility

At ORNL, scientists and engineers made excellent materials decisions, culminating in the MSRE’s demonstration of materials compatibility in a neutron environment between FLiBe salts, graphite, and Hastelloy N. The feasibility of a nuclear reactor that used LiF-BeF2 fluoride salt as a coolant with UF4 as a nuclear fuel was demonstrated in the United States in 1965. The nuclear fuel heated a coolant salt in a simple heat exchanger. The reactor had the potential for high performance because it could generate high temperatures at low pressure in a compact unit. The reactor and its primary systems were compact. Operators enjoyed the fact that the reactor was inherently safe and easy to control. Successful operation for over 20,000 hours was demonstrated. Liquid fuels enhance safety options

The reactor is equipped with a "freeze plug"—an open line where a frozen plug of salt is blocking the flow. The plug is kept frozen by an external cooling fan.

In the event of total loss of power, the freeze plug melts and the core salt drains into a passively cooled configuration where nuclear fission and meltdown are not possible. Drain tank

I Design constraints I fuel salt drain tank must be designed to be prepared to receive the entire inventory of hot fuel salt immediately after shutdown I insignificant neutron flux, strongly subcritical due to no graphite I Non-emergency use parameters I to help keep drain tank at operating temperature, plan to utilize volume for decay of noble gases I cooling requirement for noble gas decay is similar to fission-product decay heating immediately after shutdown I small amount of fuel salt held up in drain tank for decay, prior to transfer to chemical processing system I Drain tank represents the "emergency core cooling system" MSRE salt drain tank design shown, but does not represent LFTR design pixels.jpg Two-Fluid MSBR Reactor Module and Core Cutaway Coolant Choices for a Nuclear Reactor

atmospheric high-pressure pressure operation operation Metal Water

moderate temperature (250-450◦C)

Salt Gas

high temperature (650-900◦C) Fluoride salts are safe and versatile Chemically stable in air and water

Unpressurized liquid with 1000◦C range of temperature Fluorine has extremely high electronegativity, allowing fluoride salts to retain most fission products. Free energies of formation of oxides at 1000 K )

2 −50

−100

−150

−200

−250

Free Energy of Formation− (kcal/mol O 300 Th Gd Pr Nd La Ce Sr U Zr Ba Pu Cs Free energies of formation of fluorides at 1000 K )

2 −100

−150

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Free Energy of Formation− (kcal/mol F 300 Ba Sr La Gd Pr Ce Nd Cs Th Pu U Zr Free energy differences (oxides/fluorides) at 1000 K

0

−50

−100

−150 Free Energy Differences

−200 Cs Ba Sr La Pr Ce Gd Nd Pu Th U Zr In fluoride form, all actinides and alkaline fission products – most notably cesium and strontium – remain in fluoride salt form in the presence of air, and don’t form volatile species. In molten salts, the first barrier to fission product release is the chemical form of the fuel salt, rather than the mechanical integrity of the fuel pin. LiF-BeF2 fluoride salt is an excellent carrier for uranium (UF4) nuclear fuel. Element Absorption Cross-Sections

Only a few elements have neutron absorption cross sections low enough to be of interest, along with suitable chemistries.

heavies Ce Pb Bi

transitions Sn Zr

group 16 Se S O

group 15 As N P 15

group 14 Si C

B group 13 Al 11

group 17 Cl 37 F

group 2 Ca Mg Be

6 group 1 Li K NaRb HD 7

hydrogen H 2

102 101 100 10−1 10−2 10−3 10−4 10−5 Neutron Absorption Cross-Section (barn) Melting Temperatures of Salt Constituents

Individual fluoride salt components tend to have melting points too high (>600◦C) to be useful in practical reactors.

PuCl3 PuF3 plutonium UCl4 UCl3 UF4 uranium ThCl4 ThF4 thorium

PbCl2 PbF2 lead

ZrF4 zirconium RbCl RbF rubidium

CaCl2 CaF2 calcium CaCO3 KNO3 KCl KF potassium

K2CO3 MgCl2 MgF2 magnesium NaNO3 NaCl NaF sodium BeCl BeF 2 2 Na2CO3 beryllium LiNO3 LiCl Li2CO3 LiF lithium

200 300 400 500 600 700 800 900 1,000 1,100 1,200 1,300 1,400 1,500 Temperature (◦C) Lithium Fluoride Melting Temperatures

1200 CaF 2 1110 1100 NaF UF 4 1036 ThF4 996 1000 BeF2

C) 900 ◦ 848

800

chromium migrates in Hastelloy-N 700

600 Melting Temperature ( 551

500

400

300 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 Composition Neutron reactions in LiF-BeF2 salt

− 7 n,γ 8 β 8 α 4 3Li −−−−→ 3Li −−−−−→ 4Be −−−−→ 2He 0.045 b 0.842 sec 67 asec

0 9 n ,2n 8 α 4 4Be −−−−→ 4Be −−−−→ 2He 0.160 b 67 asec

− 19 n,γ 20 β 20 9F −−−−→ 9F −−−−→ 10Ne 0.009 b 1.1 sec

− 9 n,γ 10 β 10 4Be −−−−→ 4Be −−−−→ 5B 0.009 b 1.6 Myr

− 6 n,α 3 β 3 3Li −−−→ 1H −−−−→ 2He 941 b 12.3 yr

0 − 7 n ,2n 6 n,α 3 β 3 3Li −−−→ 3Li −−−→ 1H −−−−→ 2He 68 µb 941 b 12.3 yr

0 − 9 n ,α 6 n,α 3 β 3 4Be −−−−→ 3Li −−−→ 1H −−−−→ 2He 0.051 b 941 b 12.3 yr Neutron reactions in LiF-BeF2 salt

− 6 n,α 3 β 3 3Li −−−→ 1H −−−−→ 2He 941 b 12.3 yr

0 − 7 n ,2n 6 n,α 3 β 3 3Li −−−→ 3Li −−−→ 1H −−−−→ 2He 68 µb 941 b 12.3 yr

0 − 9 n ,α 6 n,α 3 β 3 4Be −−−−→ 3Li −−−→ 1H −−−−→ 2He 0.051 b 941 b 12.3 yr

Residual lithium-6 forms tritium in a neutron flux, and even if it could be eliminated, fast neutron reactions in lithium-7 and beryllium recreate lithium-6. Tritium shall be captured to the maximum degree practical at each stage in the reactor system, with the summation of these capture techniques minimizing tritium release to a degree that is acceptable from a licensing basis. Graphite undergoes dimensional change

Under sustained irradiation, the lattice structure of graphite changes. First it contracts, then it expands steadily. This effect is enhanced by an increase in temperature. Ultimately, it leads to a need for graphite replacement. To meet the challenges of liquid-fluoride salt mixtures, metallurgists at Oak Ridge developed a high-nickel alloy they called "INOR-8", but which is now more commonly known as "Hastelloy-N". Structures Fabricated from INOR-8

From this material they fabricated all the structures that would be in contact with the fluoride salt mixture. Sustainable Use of Thorium is the Ultimate Goal

recycle decay transmute sinneutrons fission

Uranium Thorium Uranium 233 232 233

Uranium-233 alone produces sufficient neutrons per thermal neutron to match or exceed its consumption. In this way, U-233 is the "catalyst" to sustainable energy production from thorium and can almost eliminate transuranic production. Fluoride Thermal Reactor Material Flows

Fluoride Thermal 1000 MW Reactor

fission thorium products

uranium- 233 Potential Steady-State Waste Profile

FP Total Inventory (MT) FP Radiotoxic Inventory (MT) 500,000 450,000 400,000 350,000 300,000 250,000 200,000 150,000 100,000 50,000 0 0 50 100 150 200 250 300 350 400 Year Conclusions

I Flibe Energy is developing the Liquid Fluoride Thorium Reactor (LFTR) I Two-fluid MSR with chemical processing, off-gas treatment and sequestration, and sCO2 PCS I FLiBe salts (HD Li), graphite channels also moderate, and thin-walled Hastelloy N vessel and piping for low pressures and corrosion I Minimal stored energy or driving forces for radionuclide release, inherent safety features I Fuel cycle potential: no mining, no enrichment, minimal fissile inventories, minimal fissile transportation, no excess fissile production, minimal fission product inventory during operation, steady-state and manageable waste stream