I have been reviewing papers, including a couple that go back to the 1970's on development issues confronting MSR/LFTR technologies before they can be useful tools for generating commercial technology. My purpose is to test a common criticism of MSR technology, namely that it is vaporware.
The first document I reviewed was WASH-1222, a document prepared during the 1970's to evaluate MSR technology. The Purpose of WASH-1222 is generally understood to have been to have the dismissal of the Molten Salt Breeder Reacto as a competitor of the ill famed LMFBR. Thus the writers of WASH-1222 were motivated to uncover flaws in the MSBR concept.
The other two documents were prepared since the beginning of the 21st century, and express views that are more favorable to the MSR/LFTR. I use the term MSR as a generic term that includes both thorium cycle and uranium cycle reactors, as well as different the use of chloride salts as well fluoride salts as fuel carrier/coolants. The Liquid Fluoride Thorium Reactor is a Thorium fuel cycle, fluoride salt, MSR.
<span style="font-weight:bold;">Where is the LFTR on the product development cycle?</span>
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:
- 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.
- 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, "<span style="font-style:italic;">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</span>".
WASH-1222 also noted, "<span style="font-style:italic;">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</span>". 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 "<span style="font-style:italic;">requirement for remote maintenance will significantly affect the ultimate design and performance of the plant system</span>". It then pointed to one of the significant problems with the MSBR design, "<span style="font-style:italic;">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</span>". 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 States government energy bureaucracy. Again the tritium problem seems no deal breaker. The ORNL researchers who were trying to solve the tritium problem stated:
"<span style="font-style:italic;">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</span>".
The ORNL researchers further summarized their findings:
<span style="font-style:italic;">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:
- The chemistry of sodium fluoroborate and the trapping process by which tritium is retained by the salt,
- Permeability values for Hastelloy N.
- Solubility data for the dissolution of elemental hydrogen (tritium) in sodium fluoroborate.
- 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.
- Identification of the sink that required saturating before steady state conditions could be established.
- 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</span>".
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.
Charles Forsberg on MSR "Developmental Gaps"
Two American scientists, Drs. Ralph Moir and Charles Forsberg have attempted to assess the state of MSR/LFTR development. (Access to papers discussed can be found here.)
In this post, I will discuss Forsberg's views on MSR/LFTR development as stated in a 2006 paper, <span style="font-weight:bold;">Molten-Salt-Reactor Technology Gaps</span> (Proceedings of ICAPP ‘06, Reno, NV USA, June 4–8, 2006, Paper 6295), in which he lays out both the case for development and discusses the developmental research required to achieve a commercial LFTR product.
Forsberg pointed to two major advantages of MSR/LFTR technology:
"<span style="font-style:italic;">As a liquid-fuel reactor, the MSR has two sets of unique characteristics relative to solid-fuel reactors.
• Safety. Under emergency conditions, the liquid fuel is drained to passively cooled critically safe dump tanks. By the use of freeze valves (cooled sections of piping) and other techniques, this safety system can be passively initiated upon overheating of the coolant salt. MSRs operate at steady-state conditions, with no change in the nuclear reactivity of the fuel as a function of time. Last, the option exists to remove fission products online and then solidify those radionuclides into a stable waste form. This minimizes the radioactive inventory (accident source term) in the reactor core and potential accident consequences.
• Fuel cycles. The liquid fuel allows online refueling and a wide choice of fuel cycle options: burning of actinides from other reactors, a once through fuel cycle, a thorium-233U breeder cycle, and a denatured thorium-233U breeder cycle. Some of the options, such as a thermal-neutron-spectrum thorium-233U breeder cycle require online refueling and thus can not be practically achieved using solid fuels. The use of a liquid fuel also avoids the need to develop fuel or fabricate fuel</span>".
Forsberg believes that recent advances in Brayton cycles gas turbines, actually solve many MSR developmental issues, and point theway to a significant improvement in MSR efficiency:
• "<span style="font-style:italic;"><span style="font-weight:bold;">Efficiency</span>. MSRs are naturally high-temperature reactors. Depending upon the choice of salt, the freezing points are between 320 and 500°C. The heat transfer properties (viscosity, thermal conductivity, etc.) improve rapidly with increasing temperature. Consequently, the detailed 1000-MW(e) conceptual design of the MSR had a reactor-core fuel-coolant exit temperature of 705°C. However, because of corrosion and other constraints in steam cycles, peak steam cycle temperatures are between 500 and 550°C. In the 1960s designs, high-
temperature heat was inefficiently dumped to lower temperatures to match what the steam cycle could tolerate. This process reduces heat exchanger sizes but has a large penalty in terms of efficiency. In contrast, many Brayton cycles operate above 1000°C. The adoption of closed helium or nitrogen Brayton power cycles enables the power cycle to efficiently use the high-temperature heat generated by the MSR. This capability allows a 15% improvement in electrical power output without changing the temperatures of the fuel salt exiting the reactor core</span>.
According to Forsberg, the choice to use Brayton cycle power generating technology would be a considerable developmental resolve these developmental issues:
• <span style="font-weight:bold;">Freeze protection</span>
• T<span style="font-weight:bold;">ritium control</span>. <span style="font-style:italic;">Adoption of a Brayton cycle provides an alternative tritium trapping option where the tritium is removed from the helium in the Brayton power cycle. This is potentially a high performance low-cost option based on demonstrated inexpensive methods to remove tritium gas or tritiated water from helium. Helium-cooled high-temperature reactors produce tritium from nuclear reactions with 3He and from leaking fuel; consequently, these reactors are equipped with systems to remove the tritium from the helium</span>.
• C<span style="font-weight:bold;">hemical reaction</span>s. "<span style="font-style:italic;">Changing from a steam cycle to a gas Brayton cycle eliminates this class of challenges.</span>"
Forsberg also reported that Pebble Bed Reactors are "<span style="font-style:italic;">being constructed in South Africa with a helium Brayton power cycle. Additional technology development would be required for an MSR; however, the closed Brayton cycle technology istransitioning to a commercial technology</span>".)
Forsberg notes second developmental shortcut:
"<span style="font-style:italic;">In the last decade, compact plate-fin and printed circuit high-temperature heat exchangers have been developed for the aircraft, chemical, and offshore-oil industries. The adoption of compact heat exchangers drastically reduces the molten fuel salt inventory in the heat exchangers and may reduce the inventory of fuel salt in the reactor by up to 50%. There are major benefits in using such heat exchangers products used in industry. These heat exchangers are being considered for use in high-temperature helium-cooled reactors and in the transport of heat from high-temperature gas-cooled reactors to hydrogen production plants using liquid-fluoride-salt heat- transport systems. Additional work is required to fully evaluate their use in MSR</span>s".
Forsberg reports the advantages of new technology heat exchanges to include:
* <span style="font-weight:bold;">Fuel salt inventory</span>. <span style="font-style:italic;">Reducing the fuel inventory reduces both fuel salt costs and nonproliferation risks, because the total fissile inventory in the nuclear system is decreased</span>.
• <span style="font-weight:bold;">Fuel salt processing</span>. "<span style="font-style:italic;">In an MSR, volatile fission products (including xenon) are removed continuously, which creates a large parasitic neutron sink in solid-fuel reactors. For nonvolatile fission products, the fuel salt is processed online or off-line, depending upon design goals. Reducing the salt inventory reduces the quantities of salt to be processed</span>".
• <span style="font-style:italic;">Heat exchanger size</span> "<span style="font-style:italic;">The size of the heat exchangers is reduced by a factor of 3 or more</span>.
• <span style="font-weight:bold;">Tritium control</span>. "<span style="font-style:italic;">The aircraft and other industries have developed compact heat exchangers with buffer gas zones to separate different fluids that may react explosively—such as hot gases vaporizing fuels in aircraft. The same technologies enable trapping of tritium from the primary system in the heat exchanger</span>".
Forsberg also addresses four further MSR/LFTR developmental issues:
* <span style="font-weight:bold;">Fuel Storage</span>
* <span style="font-weight:bold;">Noble Metal Plate-Out</span>
* High-Level Waste (HLW) Form
* <span style="font-weight:bold;">Peak Reactor Temperature</span>
The curious can read what Forsberg says about the first three by accessing Ralph Moir's MSR paper collection. I will briefly report of Forsberg's suggestions on peak reactor temperature.
According to Forsberg with current reactor construction materials, materials, the peak operating temperature is limited to around 750 degrees C. Forsberg suggests that there are a number of good reasons making higher temperature operations desirable. He states, "Carbon-carbon composites are potentially an enabling technology for very high temperature MSRs. However, there are major technical uncertainties including joining technologies". Indeed there are. The same radiation problem that effect core graphite will effect carbon-carbon composites. One possibility which Forsberg did not consider is the use of carbon nanotubes in MSR core construction. "<span style="font-style:italic;">Pound for pound, carbon nanotubes are stronger and lighter than steel</span> . . ."
<span style="font-style:italic;">Boris Yakobson, a Rice University professor of mechanical engineering and materials science and of chemistry, noted a unique feature of some carbon nanotubes, there ability to self repair. When damaged, tiny blemishes crawl over the skin of the damaged tubes, sewing up larger holes as they go. Yakobson stated, "The shape and direction of this imperfection does not change, and it never gets any larger," "We were amazed by it, but upon further study we found a good explanation. The atomic irregularity acts as a kind of safety valve, allowing the nanotube to release excess energy, in much the way that a valve allows steam to escape from a kettle</span>."
Yakobson and a research associate Feng Ding found that this mechanism had the power to heal damage to carbon nanotubes caused by radiation. Whether the nanotubes can self repair in a reactor core has yet to be tested, but carbon nanotubs have an interesting potential both as a MSR/LFTR core moderator, and as a heat and radiation resistant structural material.
Ralph Moir prepared this "<span style="font-style:italic;">partial list of research topics that could have a substantial improvement in the prospects for a commercially viable (MSR) product"</span>.
- <span style="font-weight:bold;">Thermostat negative temperature control</span>. The point of this topic is to end up with a strong negative temperature coefficient even at high temperatures and with low fission product burden. Neutron poison rods are actuated by temperature response need to develop a design that is compatible chemically and good for neutron economy and for waste management
2.<span style="font-weight:bold;"> All carbon composite primary system</span>. The point of this research idea is to be able to operate at high enough temperature (~900°C) for a direct cycle gas turbine or (~1050°C) to make hydrogen by thermo chemical water splitting cycle. We are speaking of vessel, piping, pumps and heat exchangers.
a, allows higher temperature
b. avoid corrosion of metal alloys
c. better tritium control with SiC layer
d. can SiC be used to improve the neutron damage characteristics of graphite? The problem is thermal stresses and de-bonding due to differential thermal expansion
- <span style="font-weight:bold;">Alternative salt formulation</span>s. The point of this research is to avoid the problems listed below with Li and Be.
a. Li results in tritium production and lithium-7 is expensive
b. Be is expensive and hazardous to work with due to inhalation toxicity. Look at NaF, ZrF4, look at solubility enhancement, corrosion, neutron loss.
- <span style="font-weight:bold;">Safeguard and non proliferation analyses in use of Th-U-233 cycle</span>. The point of this topic is to understand proliferation issues.
a. make most or all of fuel once started up (CR~1)
b. maybe start up on reactor minor actinides (Pu and higher); get credit for taking on this material rather than paying for enriched U
c. enhance U-232 production to promote non-proliferation by making diversion of U-233 harder, less desirable and easier to detect. See item 9 below.
- <span style="font-weight:bold;">Centrifuge for noble and semi-noble metal separations</span>. The point of this work is to improve the outlook for extraction and handling of the precipitating fission products to enhance waste management strategies.
a. base on continuous flow contactor
b. incorporate into (fuel salt) pump
- <span style="font-weight:bold;">Waste form and assay study</span>. This topic emphasizes waste management, a vital aspect of nuclear energy.
a. estimate assay (carry over of actinides) with each class of waste, e.g., gaseous, noble and semi-b. noble metals and valence two and three products with reductive extraction. Consider Bi carry over and resulting Po-210.
c. waste form: fluoride for interim and substitute fluorapatite for permanent storage
- <span style="font-weight:bold;">Plant description, size, undergrounding, cost, system</span>s. Economic considerations are important motivators to develop this new nuclear power system.
a. change out time for graphite vs. core size
b. salt formulation
c. underground design considerations
d. cost analysis preliminaries
e. power conversion cycle
- <span style="font-weight:bold;">System assessment along lines of</span> NERI-2002 <span style="font-weight:bold;">proposal</span>. What is known about the molten salt reactor is decades old. Bringing up to date the database is vital to resurrecting the molten salt reactor development.
- <span style="font-weight:bold;">U232 proliferation for thorium</span>
<span style="font-weight:bold;">A note about Ralph Moir's list</span>
Ralph Moir's list is closely related to Forsberg's analysis of MSR developmental issues. In addition Moir pushes three important developmental issues related to the integration of MSR/LFTR technology into the energy production system, the control of the cost of implementing wide scale deployment of MSR/LFTR technology, and to the prevention of nuclear weapons proliferation. We might refer to these as grand scale development issues.
If MSR/LFTR technology is to receive wide spread deployment - and I would argue that this would be our best hope for meeting the twin challenges of peak oil, and CO2 driven global warming - then economic, social, and political issues must be also addressed, as part of developmental research. I have argued in previous posts, that a "full court press" approach to cost lowering in the design, production and deployment of LFTRs could yield significant cost savings. In addition, the introduction of mass production techniques for LFTRs would lead to the ability to quickly build and deploy a very large numbers of reactors in a relatively short period of time. MSR developmental research should include research on cost lowering, mass production, site selection, and power distribution.
I personally regard the word proliferation as a shibboleth. When the words "nuclear proliferation" are uttered, morally proper people are expected to recoil in horror, and stop thinking. Actually virtual blueprints of low cost nuclear proliferation technology already exist in the public domain. Even the technologically backward, failed state of North Korea has mastered it. MSR technology would be far less attractive as a proliferation tool, than a World War II era, easily acquired, and relatively low cost technology, the Graphite Reactor. Unfortunately the value of such knowledge will be undoubtedly lost on politicians, who seldom use their intelligence, and usually lack the slightest capacity for courage. Therefore the existing proliferation resistant features of the LFTR should be assessed, and if necessary enhanced, as a part of LFTR development.