MSR/LFTR Development: WASH-1222 and Beyond
Where is the LFTR on the product development cycle?
A proof of concept MSR prototype was built in the 1950’s. It was regarded as highly successful. An advanced MSRE prototype was built and tested between 1966 and 1969. It was, like the first prototype, considered an outstanding successes. The MSRE Accomplished all experimental objectives The MSRE, tested many advanced technologies, including
* Online reactor refueling
* First reactor to use U-235, U-233, and Pu-239 as nuclear fuel
* The longest reactor runs between shutdowns at the time
* Verified MSR safety features
* Successfully use of the liquid LiF-BeF2-ZrF4-UF4 fuel/coolant formula.
Several developmental problems emerged from the test:
1. Tritium, a radioactive form of hydrogen, was noted to have escaped the reactor. This was highly undesirable, but not entirely unexpected. Thus a tritium control system had to be developed. This was later accomplished.
2. Cracking on the surface of metals that came in contact with liquid salts was observed. Later research found the cause of the of the cracking and a method of preventing it.
3 Prolonged and heavy neutron radiation exposure of graphite, lead to changes of graphite internal structure. This produced swelling of the graphite moderators which also served as the inner plumbing of the reactor. The changes in graphite structure weakened it. This problem has yet to be solved, but can be worked around.
Upon completion of the MSRE, ORNL staff began to design a large (1000 MWe) LFTR, the MSBR. It was designed to serve as a thorium fuel cycle breeder. The jump from the 10MW MSRE to a 3000 MW MSBR was in hindsight overly ambitious, but was necessitated by AEC requirements. The solution to the graphite problem was particularly unsatisfactory. Core graphite was to be periodically removed and replaced. There are several less drastic alternatives.
Wash-1222 listed a number of developmental issues facing the MSBR design and development team. Wash-1222 stated, “the development of these larger components along with their special handling and maintenance equipment is probably one of the most difficult and costly phases of MSBR development. However, reliable, safe, and maintainable components would need to be developed in order for any reactor system to be a success“.
WASH-1222 also noted, “The salt valves for large MSBR’s represent another development problem, although the freeze valve concept which was employed successfully in the MSRE could likely be scaled up in size and utilized for many MSBR applications. Mechanical throttling valves would also be needed for the MSBR salt systems, even though no throttling valve was used with the MSRE. Mechanical shutoff valves for salt systems, if required, would have to be developed“. This would seem to be a simple developmental task.
WASH-1222 also noted that an integrated fuel reprocessing system would have to be tested, and a design for system integration for the entire MSBR was also required.
Many of the developmental tasks listed by WASH-1222 apply primarily or entirely to the MSBR. Other developmental tasks appeared to be routine and not likely to pose a challenge.
WASH also noted the MSBR “requirement for remote maintenance will significantly affect the ultimate design and performance of the plant system“. It then pointed to one of the significant problems with the MSBR design, “the removal and replacement of core internals, such as graphite, might pose difficult maintenance problems because of the high radiation levels involved and the contamination protection which would be required whenever the primary system is opened“. This pointed to the most significant problem of the MSBR design, the resolution of the graphite problem by periodic core removal. French MSR researchers, have recently made the choice to follow a developmental track that eliminates graphite from the core of their proposed MSR. There analysis of the difficulties posed by the graphite core of the MSRE, lead them to conclude that despite some significant disadvantages, the a graphite free core offered more advantages.
WASH-1222 raised questions about the safety of the MSBR. Subsequent MSR safety analysis by Uri Gat, and Gat and Dodds, would seem to resolve most safety questions on a conceptual level. Recent discussions in the “Energy from Thorium” raised questions about assurances that the “salt freeze safety valve would operated in a timely fashion in the event of an emergency shut down. My rather brief review of ORNL reports did not shed light on the question. In absence of devinitive evidence from ORNL reports, the proper functioning of the emergency reactor drain system including the freeze valve, should be verified, and any short comings rectified.
Thus the major MSBR developmental problems noted by WASH-1222 were the tritium problem, and the problem of core graphite. The tritium problem requires a technological fix that is clearly not impossible. Several work around ideas have been proposed for the graphite problem, and a French MSR design team has adopted one.
In addition to the developmental issues noted by WASH-1222, the problem of protactinium extraction, a problem that bedeviled my father from the late 1950’s to the mid 1960’s, has been the subject of continuing discussions on “Energy from Thorium”. The tennor of the discussion seems to be as follows, protactinium extraction is difficult and probably should be avoided if possible.
I mentioned alternative approaches to the graphite problem. Again some available options have been discussed on “Energy from Thorium”. These include the big pot approach which has attracted french interest. The reactor core is simply a open chamber into which liquid salt coolant/fuel is poured. No moderator is used although the liquid coolant/fuel does have some moderating effect. There are disadvantages to this approach. The amount of fissionable fuel required to sustain a chain reaction would be much greater that in a moderated MSR.
One interesting option would be to put graphite pebbles into the pot in order to provide a moderator. The graphite pebbles would float in the liquid salt and could be periodically removed for replacement. This system was actually suggested at ORNL in 1970.
“Jaro” suggested the use of self-cleansing carbon nanotubes as MSR moderators. Another “jaro” suggestion involved the use of heavy water being piped through the MSR core. There would probably be safety concerns about this design, although heavy water would work even better as a moderator that graphite.
It would appear then that the graphite problem is no deal killer for the MSR. Solutions and work arounds exist for the graphite problem, but reactor developers have to decide which one to choose.
Finally, research on the tritium problem was problem was continued at ORNL into the mid 1970’s. Tritium (H-3) is a radioactive isotope of hydrogen that primarily is produced from lithium-6 isotopes. If pure lithium-7 is used in the fuel, then the LFTR tritium problem would be greatly reduced, but not entirely eliminated. Tritium hike the other forms of hydrogen diffuse through metal barriers. Tritium is most likely to escape the MSR/LFTR through the thin walls the heat exchange. ORNL researchers in 1977 later reported that they were making progress toward a solution to the tritium problem when their funding was cut off by the United Sta
tes government energy bureaucracy. Again the tritium problem seems no deal breaker. The ORNL researchers who were trying to solve the tritium problem stated:
“Although a complete understanding of the behavior of tritium in sodium fluoroborate could not be developed from this series of experiments due to the termination of the Molten-Salt Reactor- Program, the effectiveness of sodium fluoroborate to trap tritium was demonstrated. Furthermore, use of sodium fluoroborate as a secondary coolant in an MSBR would be expected t:o adequately limit the transport of tritium to the reactor steam system and environment“.
The ORNL researchers further summarized their findings:
The tritium addition experiments conducted in the CSTF demonstrated sodium fluoroborate’s effectiveness for sequestering tritium. However, further experimentation and research would be required to yield a better understanding of tritium behavior in sodium fluoroborate, to better define
basic parameters, and to explain some of the observed phenomena as a result
of conducting the experiments in the CSTF.
If the MSR program were to be continued, further investigation relating to the following would be desirable:
1. The chemistry of sodium fluoroborate and the trapping process by which tritium is retained by the salt,
2. Permeability values for Hastelloy N.
3. Solubility data for the dissolution of elemental hydrogen (tritium) in sodium fluoroborate.
4. Data on gas-liquid equilibria in the pump bowl in an effort to
explain behavior such as that observed in experiment T4 when, upon increasing the off-gas flow rate to 4 liters/min, equilibrium conditions in the pump bowl between the gas and liquid were altered drastically.
5. Identification of the sink that required saturating before steady state conditions could be established.
6. Determination of the existence of an extraneous source of hydrogen in the off-gas system and its effect (if present) on the behavior and distribution of tritium in the CSTF“.
Thus the obstacles to successful development of the MSR/LFTR mentioned by the WASH-1222, can be. Design choices and promising research avenues known since the 1970’s are still available.