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Office of Nuclear Energy

Molten Salt Fuel Definition, Specification, and Qualification

GAIN Advanced Fuels WorkshopMarch 5-6, 2019, Boise, ID

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• Nuclear fuel: Discretely engineered elements or liquids containing fissile nuclear material that have the thermophysical, thermochemical, material, and geometric properties for sustaining criticality within the nuclear reactor for which it was designed.

• Fuel qualification: A process which provides high confidence that physical and chemical behavior of fuel is sufficiently understood so that it can be adequately modeled for both normal and accident conditions, reflecting the role of the fuel design in the overall safety of the facility. Uncertainties are understood such that any calculated fission product releases include appropriate margin to ensure conservative calculation of radiological dose consequences.

Reasonable definitions

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• MSR program did experimentally and computationally evaluate fuel salt chemical and physical behavior under both normal and accident conditions• Experimentally demonstrated fuel salt radionuclide release amounts

• Relied on low-pressure, multi-layered containment, negative reactivity feedback, and effective semi-passive decay heat removal to ensure minimal radiological dose consequences

Fuel Qualification Did Not Exist as a Formal Concept at the Time of the MSRE

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1. Demonstrated radiolytic stability2. Evaluated radionuclide release amounts3. Evaluated material compatibility• Fuel salt capsules irradiated to assess volatility of fission

products (particularly iodine, tellurium, and ruthenium)• Two sealed capsules• Two connected to tubing for on-line gas analysis• Separate purge-gas lines and sampling lines

• In-pile natural circulation fuel salt loop operated to assess fission product behavior• Loop included free surface continuously purged with dry helium• Following operation loop was sectioned and fission product

activities measured throughout

MSR Program Featured Multiple Fuel Salt Irradiation Experiments

ORNL-4170ORR In-Pile Loop #2

Typical Capsule Assembly Equipped for

Purging, Experiment ORNL-MTR-47-6

ORNL-3789

4Capsule tests performed in thermal neutron fluxes from 1011 to 1014 neutrons/cm2-sec and power levels from 80 to 8000 w/cm3 — Fluid Fuel Reactors 1958

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Potential For Radiation Acceleration of Corrosion Was Experimentally Evaluated

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The effects of irradiation on the corrosion of Inconel exposed to fluoride fuel mixtures and on the physical and chemical stability of the fuel mixtures have been investigated by irradiating in the MTR capsules filled with static fuel and by operating in-pile forced-circulation loops in the LITR and in the MTR. In the many capsule tests and in the three in-pile loop tests made to date, no major changes have occurred in the fuel mixtures that can be attributed to irradiation, other than normal burn-up of uranium. Metallurgical examinations of the Inconel capsules and tubing have likewise shown no changes in corrosion that can be the result of radiation damage. – Kielholtz et al. -1959 LITR Fuel Salt Test Loop Diagram

ORNL-1965

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• Actinide solubility• Fast spectrum fluoride salts operate near solubility limits• Log of actinide trifluoride solubility is roughly linear

versus inverse temperature• Vapor pressure

• Relative volatility of rare earth fission products measured – Hightower et. al - 1968

• Vapor pressure of UCl4 at 800 °C is ~900 mm. Hg. Over fuel melt with UCl4/UCl3 = 0.1 actually ~0.7 mm. Hg –source Harder et al. 1969

• Oxygen contamination• Oxygen substantially increases oxidative corrosion• ZrF4 included in MSRE fuel salt to precipitate as ZrO2

upon oxygen contamination instead of UO2

ZrO2(s) + UF4 (d) ⇄ ZrF4(d) + UO2(s) driven leftward by ZrF4

• Chloride salts rely on adequate solubility of uranium oxides

Fuel Salt Chemistry Was Initially Evaluated Under Non-Radiation Conditions

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C. J. Barton, “Solubility of Plutonium Trifluoride in Fused-Alkali Fluoride-Beryllium Fluoride Mixtures” J. Phys. Chem. , Vol. 64, 1960

Molten Salt Still

ORNL-TM-2058

Solubility of PuF3 in FLiBe

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W. D. Powers, S. I. Cohen & N. D. Greene (1963) Physical Properties of Molten Reactor Fuels and Coolants, Nuclear Science and Engineering, 17:2, 200-211

• Viscosity – ORNL employed capillary, oscillating cup, rotational, and falling ball viscometers

• Density – Archimedes or dilatometer• Heat capacity – ice or copper block calorimeter• Thermal conductivity – variable gap technique

Extensive Molten Salt Physical Property Measurements Performed

7Dilatometer for molten salts

ORNL-TM-4308ORNL-1040

Bunsen Ice Calorimeter

Variable Gap Thermal Conductivity

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• NRC has tentatively agreed with proposed fuel qualification approach subject to agreement by prospective applicants• NRC is soliciting comments on this report from NEI MSR

technical working group• Approaches focuses on fuel salt thermochemical and

thermophysical properties• Analogous to materials surveillance program• Requires baseline property measurements and property

evolution prediction models• Based upon bounding values of fuel salt properties

necessary for adequate performance during anticipated operational occurrences and design basis accidents

• Established through accident progression models

ORNL Recently Proposed a Predict and Verify Based Approach to MSR Fuel Qualification

8https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML18347A303

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• Validated fuel salt property model• Current campaign activity

• Validated baseline fuel salt property measurements• Current campaign activity

• Validated measurement techniques that can be employed at plant sites• Design basis accident set

• Licensing basis event workshop planned• Accident progression modeling tools

• Radionuclide releases (especially off gas system)

Additional Information Is Needed for Fuel Salt Qualification

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