[I]t is incumbent on those in high positions to reach wise decisions, and it is reasonable and important that the public be correctly informed. It is incumbent on all of us to state the facts as forthrightly as possible. – Hyman Rickover testifying before Congress in 1953,
QUESTION FROM SENATOR SHAHEEN
Q3. Of the six Gen IV nuclear power technologies proposed by the US in 2000, DOE Idaho National Labs have been pursuing two – (1) high temperature gas-cooled reactors for hydrogen production, and (2) sodium-cooled fast reactors for waste burning. Separately, liquid-fluoride thorium reactor research is ongoing at UC Berkeley, MIT, Redstone Arsenal, and in other countries including France, Japan, and Canada.
As the Department analyzes advanced reactor designs, can you tell me if the liquid-fluoride thorium reactors are under consideration? What are the benefits of liquid-fluoride thorium reactors? What are the drawbacks or downsides of liquid-fluoride thorium reactors? How does power generated from liquid-fluoride thorium reactors compare, on a price per kilowatt hour, with power generated from the current coal generation fleet in the United States? As we confront our nation’s energy and climate challenges, what role might these types of reactors play?
A3: The “liquid-fluoride thorium reactor,” otherwise known as a molten salt reactor (MSR), where molten salts containing fissile material circulate through the reactor core, is not part of the Office of Nuclear Energy’s research program at this time. Some potential features of a MSR include smaller reactor size relative to light water reactors due to the higher heat removal capabilities of the molten salts and the ability to simplify the fuel manufacturing process, since the fuel would be dissolved in the molten salt. One significant drawback of the MSR technology is the corrosive effect of the molten salts on the structural materials used in the reactor vessel and heat exchangers; this issue results in the need to develop advanced orrosion-resistant structural materials and enhanced reactor coolant chemistry control systems. In addition, operational practices would have to address the fact that the liquid salts solidify between temperatures of 300 C to 500 C, thereby requiring the use of special heating systems when the reactor is not operating. From a non-proliferation standpoint, thorium-fueled reactors present a unique set of challenges because they convert thorium-232 into uranium-233 which is nearly as efficient as plutonium-239 as a weapons material. A cost per kilowatt hour estimate has not been developed.
From ORNL/TM-6002 (J. R. Keiser, 1977):
As a result of these studies, we have found that Hastelloy N exposed in salt containing metal tellurides such as Li Te and Cr Te undergoes grain boundary embrittlement like that observed in the MSRE. The embrittlement is a function of the chemical activity of tellurium associated with the telluride. The degree of embrittlement can be reduced by alloying additions to the Hastelloy N. The addition of 1 to 2 % Nb significantly reduces embrittlement, but small additions of titanium or additions of up to 15% Cr do not affect embrittlement. We have found that if the U(IV)/U(III) ratio in fuel salt is kept below about 60, embrittlement is essentially prevented when CrTel.266 is used as the source of tellurium.
From ORNL/TM-6415 (1979):
The nickel-based alloy Hastelloy N, which was specifically developed for use in molten-salt systems, was used in construction of the MSRE.
The material generally performed very well, but two deficiencies became
apparent: (1) the alloy was embrittled at elevated temperatures by ex-
posure to thermal neutrons and (2) it was subject to intergranular sur-
face cracking when exposed to fuel salt containing fission products.
Recent development work indicates that solutions are available for both
these problems. Details of this work are given by McCoy; a summary of
the results follows
Irradiation experiments early in the MSR development program showed
that Hastelloy N was subject to high-temperature embrittlement by thermal
neutrons. The MSRE was designed around this limitation (stresses were
low and strain limits were not exceeded), but the development of an im-
proved alloy became a prime objective of the materials program. It was
found that a modified Hastelloy N containing 2% titanium had much im-
proved postirradiation ductility, and extensive testing of the new alloy
W ~ S under way at the close sf MSRE operations.
The second problem, intergranular surface cracking, was discovered
at the close of the MSRE operation when surface cracks were observed
after strain testing of Hastelloy K specimens that had been exposed to
fuel salt. Research since that time has shown that this phenomenon is
the result of attack by tellurium, a fission product in irradiated fuel
salt, on the grain boundaries.
As a result of research from 1974 to 1976, two likely solutions to the problem of tellurium attack have been developed. The first involves the development of an alloy that is resistant to tellurium attack but still retains the other required properties. This development has proceeded sufficiently to show that a modified Hastelloy N containing about 1% niobium has good resistance to tellurium attack and adequate resistance to thermal-neutron embrittlement at temperatures up to 650°C. It was also found that alloys containing titanium, with or without niobium, exhibited superior neutron resistance but were not resistant to tellurium attack.
The second likely solution involves the chemistry of the fuel salt.
Recent experiments indicate that intergranular attack on Hastelloy N
is much less severe when the fuel-salt oxidation potential, as measured
by the ratio of U4+ to U3+, is less than 60, the possibility that the superior titanium-modified Hastelloy N could This discovery opens up be used for MSRs through careful control of the oxidation state of the Fuel salt.
Bath of the above solutions appear promising, but extensive testing
under reactor conditions would be required before either could be used
in the design of a future MSR.
Also from From ORNL/TM-6415 (1979)
SPECIAL DEVELOPMENT REQUIREMENTS FOR THE DMSR
Recent reexamination of the MSR concept with special attention to
antiproliferation considerations has led to the identification of two
preliminary design concepts for MSRs that appear to have substantially
less proliferation sensitivity without incurring unacceptable performance penalties. tor) has been applied to both of these concepts because each would be fueled initially with 235U enriched to no more than 20% and would be operated throughout its lifetime with denatured uranium. The designation DMSR (for denatured molten-salt reactor) has been applied to both of these concepts because each would be fueled initially with 235U enriched to no more than 20% and would be operated throughout its lifetime with denatured uranium.
The simpler of these DMSR concepts6 would completely eliminate on-line chemical processing of the fuel salt for removal of fission products. (Stripping of gaseous fission products would be retained, and batch-wise treatment to control oxide contamination probably would be required.) This reactor would require rautine additions of denatured 235U fuel, but would not require replacement or removal of the in-plant inventory except at the end of the 30-year plant lifetime. Adding an on-line chemical processing facility to the 30-year9 once-through reactor provides the second DMSR design concept. With this addition, the conversion ratio of the reactor would reach 1.0 (i.e., break-even breeding) so that fuel additions could be elim
inated and a given fuel charge could be used in-definitely by transferring it to a new reactor plant: at the decomissioning of the old unit.
The required chemical processing facility for a DMSR, shown as a preliminary conceptual flowsheet in Pig. S.1, would be derived largely from the MSBR but would contain some significant differences. In particular, isolation and segregation of protactinium would be avoided, provisions would be made to retain and use the plutonium produced from 238U and a special step would be added for removal of fission-product zirconium. Thus, the development of on-line chemical processing for a DMSR would require essentially all the technology development identified for the
MSBR with additions to accommodate these differences. However, since the DMSR offers a no-processing option, a large fraction of the reprocessing development, along with its associated materials development, could be deferred or even eliminated to reduce the cost (but probably not the time) for developing the first DMSRs. To provide an overall perspective, this development plan includes costs and schedules for developing the reprocessing capability in parallel with the reactor. Such deferral might be expected