It seemed like a great future, the nuclear age. The promise of unlimited energy and no pollution but is that what we got out of all this?
After all the military influence and investment did we get anything good for civilians? Perhaps yes in medicine and chemistry. Besides the %20 of our energy coming from nuclear, that is now being adsorbed by higher efficiencies and solar/wind renewable sources, in the last 60 years we have had numerous problems that have killed, injured, and costed taxpayers uncontrolled losses.
|Date||Location of accident||Description of accident or incident||Dead||Cost
|September 29, 1957||Mayak, Kyshtym, Russia||The Kyshtym disaster was a radiation contamination incident that occurred at Mayak, a Nuclear fuel reprocessing plant in the Soviet Union.||6|
|July 26, 1957||Simi Valley, California, United States||Partial core meltdown at Santa Susana Field Laboratory’s Sodium Reactor Experiment.||0||32|
|October 10, 1957||Sellafield aka Windscale fire, Cumberland, United Kingdom||A fire at the British atomic bomb project destroyed the core and released an estimated 740 terabecquerels of iodine-131 into the environment. A rudimentary smoke filter constructed over the main outlet chimney successfully prevented a far worse radiation leak and ensured minimal damage.||0||5|
|January 3, 1961||Idaho Falls, Idaho, United States||Explosion at SL-1 prototype at the National Reactor Testing Station. All 3 operators were killed when a control rod was removed too far.||3||22||4|
|October 5, 1966||Frenchtown Charter Township, Michigan, United States||Partial core meltdown of the Fermi 1 Reactor at the Enrico Fermi Nuclear Generating Station. No radiation leakage into the environment.||0||132|
|January 21, 1969||Lucens reactor, Vaud, Switzerland||On January 21, 1969, it suffered a loss-of-coolant accident, leading to a partial core meltdown and massive radioactive contamination of the cavern, which was then sealed.||0||5|
|1975||Sosnovyi Bor, Leningrad Oblast, Russia||There was reportedly a partial nuclear meltdown in Leningrad nuclear power plant reactor unit 1.|
|December 7, 1975||Greifswald, East Germany||Electrical error causes fire in the main trough that destroys control lines and five main coolant pumps||0||443||3|
|January 5, 1976||Jaslovské Bohunice, Czechoslovakia||Malfunction during fuel replacement. Fuel rod ejected from reactor into the reactor hall by coolant (CO2).||2||4|
|February 22, 1977||Jaslovské Bohunice, Czechoslovakia||Severe corrosion of reactor and release of radioactivity into the plant area, necessitating total decommission||0||1,700||4|
|March 28, 1979||Three Mile Island, Pennsylvania, United States||Loss of coolant and partial core meltdown due to operator errors. There is a small release of radioactive gases. See also Three Mile Island accident health effects.||0||2,400||5|
|September 15, 1984||Athens, Alabama, United States||Safety violations, operator error, and design problems force a six-year outage at Browns Ferry Unit 2.||0||110|
|March 9, 1985||Athens, Alabama, United States||Instrumentation systems malfunction during startup, which led to suspension of operations at all three Browns Ferry Units||0||1,830|
|April 11, 1986||Plymouth, Massachusetts, United States||Recurring equipment problems force emergency shutdown of Boston Edison’s Pilgrim Nuclear Power Plant||0||1,001|
|April 26, 1986||Chernobyl, Chernobyl Raion (Now Ivankiv Raion), Kiev Oblast, Ukraininan SSR||Overheating, steam explosion, fire, and meltdown, necessitating the evacuation of 300,000 people from Chernobyl and dispersing radioactive material across Europe (see Effects of the Chernobyl disaster)||30 direct, 19 not entirely related and 15 minors due to thyroid cancer, as of 2008.||6,700||7|
|May 4, 1986||Hamm-Uentrop, West Germany||Experimental THTR-300 reactor releases small amounts of fission products (0.1 GBq Co-60, Cs-137, Pa-233) to surrounding area||0||267|
|March 31, 1987||Delta, Pennsylvania, United States||Peach Bottom units 2 and 3 shutdown due to cooling malfunctions and unexplained equipment problems||0||400|
|December 19, 1987||Lycoming, New York, United States||Malfunctions force Niagara Mohawk Power Corporation to shut down Nine Mile Point Unit 1||0||150|
|March 17, 1989||Lusby, Maryland, United States||Inspections at Calvert Cliff Units 1 and 2 reveal cracks at pressurized heater sleeves, forcing extended shutdowns||0||120|
|March 1992||Sosnovyi Bor, Leningrad Oblast, Russia||An accident at the Sosnovy Bor nuclear plant leaked radioactive gases and iodine into the air through a ruptured fuel channel.|
|February 20, 1996||Waterford, Connecticut, United States||Leaking valve forces shutdown Millstone Nuclear Power Plant Units 1 and 2, multiple equipment failures found||0||254|
|September 2, 1996||Crystal River, Florida, United States||Balance-of-plant equipment malfunction forces shutdown and extensive repairs at Crystal River Unit 3||0||384|
|September 30, 1999||Ibaraki Prefecture, Japan||Tokaimura nuclear accident killed two workers, and exposed one more to radiation levels above permissible limits.||2||54||4|
|February 16, 2002||Oak Harbor, Ohio, United States||Severe corrosion of control rod forces 24-month outage of Davis-Besse reactor||0||143||3|
|August 9, 2004||Fukui Prefecture, Japan||Steam explosion at Mihama Nuclear Power Plant kills 4 workers and injures 7 more||4||9||1|
|July 25, 2006||Forsmark, Sweden||An electrical fault at Forsmark Nuclear Power Plant caused one reactor to be shut down||0||100||2|
|March 12, 2011||Fukushima, Japan||A tsunami flooded and damaged the plant’s 5 active reactors, drowning two workers. Loss of backup electrical power led to overheating, meltdowns, and evacuations. One man died suddenly while carrying equipment during the clean-up. The plant’s 6th reactor was inactive at the time.||2+||7|
|12 September 2011||Marcoule, France||One person was killed and four injured, one seriously, in a blast at the Marcoule Nuclear Site. The explosion took place in a furnace used to melt metallic waste.||1
The better choice for nuclear power is the LFTR design.
The liquid-fluoride thorium reactor concept has strong safety advantages over today’s nuclear reactors and the potential to implement a highly efficient and sustainable fuel cycle. It can potentially produce valuable products in addition to electrical energy that will enhance its competitiveness relative to low-cost natural gas and petroleum. The capital cost structure of a LFTR would make it attractive to regulated electrical utilities that desire to maximize shareholder return while providing low-cost electrical energy to the ratepayers in their service area.
Two-Fluid MSBR Core Designs
The earliest designs for thermal-spectrum molten-salt breeder reactors using thorium were conceived before graphite was known to be a suitable moderator material. They primarily relied on the moderating properties of the salt itself, which were poorer than graphite. Because of this limitation they could not achieve a truly thermal-neutron spectrum and this led to higher fuel inventories. The absorption of neutrons in the salt constituents (lithium, beryllium, and fluorine) is rather low, but not nearly so low as graphite, so there were larger neutronic losses to the salt itself.
The barrier material between the core and blanket salts was metallic and this structure would have a limited lifetime, if it even was feasible.
In-pile experiments with graphite in the early 1960s established its chemical and neutronic stability with the fuel and blanket salts, allowing graphite-moderated reactors to begin to be considered in future designs. The first fruit of this realization was the decision to manufacture unclad graphite moderator elements for the Molten-Salt Reactor Experiment, which had been approved for construction in 1960.
ORNL-3708, February 1964-July 1964
In a 1964 progress report (ORNL-3708), the researchers of the MSRP put forward a concept for a thorium breeder reactor that first used a prismatic core design, where the prismatic structures consisted of graphite extrusions. Their primary motivation for attempting the challenging goal of the breeder was the unforgiving nature of resource calculations for reactors that fell much short of breeding.
Realization of a system that makes full use of the potential energy in thorium to produce cheap electricity is the primary mission of reactor development at the Oak Ridge National Laboratory. That system must be an efficient breeder system. An advanced converter may be a worthwhile step in the development, but an advanced converter does not reach the goal. No matter how good the conversion ratio, if it is significantly less than 1, the amount of uranium that must be mined to make up the deficit in fissionable material is greater than the amount of thorium that must be mined to compensate for the thorium converted to 233U and burned. For example, if the conversion ratio is 0.90, the 235U from 20 tons of natural uranium will be burned with each ton of thorium consumed. Even with a conversion ratio of 0.99, the 235U from 2 tons of uranium must be supplied with each ton of thorium.
Their concept for the breeder reactor, shown in Figure 2, used the graphite prisms as the flow channels for the fuel salt. Each graphite prisms had a square cross-section at the bottom and top of the structure, which was shaped into a circular cross-section in the “core” region between the ends. A central circular channel was bored into each graphite prism. Blanket salt was allowed to flow into interstitial regions between the graphite blocks. At the top and bottom of each set of graphite blocks, flexible piping redirected the upcoming flow from one block into the downgoing flow into an adjacent block. The graphite blocks were arranged on a square pitch (distance between centers) of 8 inches. The complication of that arrangement can be seen in Figure 3.
The overall average core power density was 40 kW/liter and the core consisted of 324 prismatic elements. Fuel salt entered the reactor at 1125°F and left at 1300°F, for a ΔT across the reactor of 175°F. The proposed coolant was a fluoride salt mixture (LiF-NaF-KF) that would enter the primary heat exchanger at 950°F and exit at 1100°F. This led to an anticipated thermal efficiency of 42-45%. The design power of the reactor was 1000 MWe from a core power of 2400 MWt.
The reactor vessel was to be constructed of a modified Hastelloy-N alloy and would be 240 inches in outer diameter. A graphite reflector 12 inches thick lined the interior of the reactor vessel. No special structure was described to achieve a transition from the prismatic structure of the core to the annular structure of the reflector and core vessel. There was simply a gap between the prismatic core and the reflector that would be filled by blanket fluid, as can be seen in Figure 2.
This design proposed to process the reactor’s fuel inventory in 10-60 days and the blanket inventory every 35-200 days. The anticipated breeding ratio was 1.08, but this had not been rigorously verified. Based on a fissile inventory of 880-1220 kg, the reactor achieved a specific fissile inventory of 0.88 to 1.22 kg of fissile for each electrical megawatt generated.
ORNL-3936, September 1965-February 1966
Studies of a thorium breeder reactor began in earnest in September 1965, following the successful completion and startup of the Molten-Salt Reactor Experiment. In the MSRP semiannual report of February 1966 (ORNL-3936), a new reactor core design was put forward that showed a number of design innovations. The reactor was sized for 1000 MWe from 2220 MWt, assuming a nearly 45% efficient steam turbine power conversion system and subtracting for station loads. The average core power density had been doubled to 80 kW/liter as well, which led to a more compact reactor vessel 120 inches in outer diameter and 150 inches high, shown in Figure 4.
The reactor core was also prismatic, but now the up-and-down flow of fuel salt through the core would be confined to a single “fuel” cell rather than passing from one channel to another, as shown in Figure 5. The graphite prisms were hexagonal in cross-section rather than square and set on a triangular pitch of 4.8 inches, with 534 in total. Salt flowed upward through the prism through eight small circular channels, and then was turned at the top of the prism by a plug structure and flowed downward through a central circular bore. This strategy eliminated the external plumbing of the ORNL-3708 core module design, internalizing the flow into a single graphite prism. Blanket salt filled the interstices between the prismatic structures, and the nominal core composition was 75% graphite, 18% fuel salt, and 7% blanket salt by volume.
The prismatic core structure was set into the reactor vessel mounted to the top of a plenum structure. Inside the plenum structure the exit plenum was fully enclosed in the entrance plenum. Each “fuel cell” was connected to each plenum by graphite-to-metal transition sleeves. The fuel cells were only anchored at one end to permit axial movement due to thermal expansion. There was no transition region between the prismatic core and the annular 3-inch graphite reflector immediately inside the reactor vessel. This can be seen in Figures 4 and 6.
Fuel salt entered the reactor at 1000°F and left at 1300°F, for a ΔT of 300°F. A mixture of sodium fluoroborate and sodium fluoride (39-61 mole%) was proposed that would enter the primary heat exchanger at 850°F and exit at 1125°F. The blanket salt was held to a smaller temperature range, entering the reactor at 1150°F and leaving at 1250°F (ΔT = 100°F).
Fuel would be processed every 47 days and blanket every 23 days. The total fuel volume in the primary circuit and processing system was 717 ft3 and the total blanket volume was 3383 ft3. With the high core power density this design projected a fissile inventory of only 769 kg, leading to a specific fissile inventory of 0.77 kg/MWe. The high core power density diminished the breeding ratio to 1.049, most likely due to neutron loss to protactinium. The fertile inventory of 260,000 kg of thorium was over 300 times greater than the fissile.
ORNL-4119, September 1966-February 1967
The next semiannual report in August 1966 (ORNL-4037) proposed a substantial change to the design approach, employing four 250-MWe modules rather than a single 1000-MWe reactor. There was also discussion of how the removal of protactinium in the blanket (rather than waiting for the protactinium to decay to uranium) would improve overall performance.
The February 1967 semiannual report (ORNL-4119) adopted the modular 250-MWe reactors (each with 556 MW of thermal power) as the baseline approach. The average power density of the core was reduced to 40 kW/liter. The graphite “fuel cell” that made up the primary structure of the core was modified from a hexagonal cross-section to a circular cross-section, such that the fuel cell was essentially a long graphite cylinder. The fuel cells were enlarged to 5 inches in diameter and their number reduced to 336. In their center was a circular bore 1.5 inches in diameter, surrounded by three 7/8-in diameter holes for the upward flow from the entrance plenum, as shown in Figure 7.
Temperature ranges for fuel, blanket, and coolant were all retained from the previous design. The plenum structures evolved but remained similar in concept, with a central feed and drain rather than an offset design. A transition structure was first proposed in this design, employing graphite spheres 4 inches in diameter to fill the region between the prismatic core and the annular graphite reflector. Each graphite ball had holes in them so that blanket salt occupied 60% of the volume of the transition region. The reflector’s thickness was increased to 6 inches. These changes can be seen in Figure 8.
Fuel and blanket processing times remained largely the same but breeding ratio improved to 1.07 due to the lower core power density of the design.
A control rod was envisioned for the central position in the core. The control rod would not be a rod in the conventional sense, but there would be a hollow graphite cylinder 5 inches in diameter, the same as the other fuel cells, and equipped with a gas inlet at the top. The hollow graphite cylinder would naturally fill with blanket salt, which was a neutron absorber and would be particularly effective in this central region of the reactor. Gas pressure would be used to position the height of the blanket salt column at any position desired within this cell, and this column height would be used for control of the reactor. If the gas were expelled, the column would fill with blanket salt, introducing negative reactivity and tending towards reactor shutdown.
An alternative approach would be to position a graphite rod 5 inches or less in diameter in this central position and to actuate it from above. By pushing the rod down into the blanket, positive reactivity would be introduced, and by withdrawing the rod and allowing the volume to fill with blanket fluid, negative reactivity would be introduced. Since the blanket salt was more dense than graphite, a loss of actuation capability would cause the graphite rod to float upward in the reactor, ideally leading to a reduction in reactivity.
ORNL-4191, March 1967-August 1967
The next semiannual report from August 1967 (ORNL-4191) was the last to feature the two-fluid design. Because design work shifted to the one-fluid core design, the two-fluid design described in ORNL-4191 became the “reference” two-fluid design, and was written up in greater detail in the final report on two-fluid reactor design efforts, ORNL-4528, Two-Fluid Molten-Salt Breeder Reactor Design Study, published in August 1970.
The “reference” two-fluid design described in ORNL-4191 and ORNL-4528 reduced the core power density further, from 40 kW/liter to 20 kW/liter, in order to improve graphite lifetime in the core. There were also significant changes to the graphite “fuel cells”. The geometry was again changed back to a hexagonal cross-section, with an inner diameter (flat-to-flat distance) of 5.375 inches. A single circular channel 2-23/32 inches in diameter was to be bored down the centerline of each graphite prism. Inside that channel would be another cylindrical channel of graphite 3/4-inch in thickness and 2-1/4 in diameter. Salt would flow upward on the outside of this “recursive” channel and then flow downward on the inside, as shown in Figure 9. With this technique, the need to drill multiple passages down the length of graphite fuel cells was done away with, and this could lead to a substantial improvement in manufacturability. The concentric flow channels were generated by placing the inner cylindrical graphite structure in the bore of the larger hexagonal graphite prism. This simple approach to a “recursive” fuel channel has been mimicked in our newer designs.
The prismatic pattern of core structures continued across the entire cross section of the core, but comprised three distinct regions. The three regions can be seen in the core cutaway view shown in Figure 12. The central region of the reactor was composed of the hexagonal “fuel cells”, in which the fuel salt flowed upward through a central bore and then downward through a concentric graphite sleeve inside the bore. There were to be 420 of these structures in the reactor vessel. Blanket salts fills the interstitial space between graphite fuel cells. Surrounding the “core” of hexagonal graphite prisms is a “blanket” constructed of simple graphite cylinders, open at both ends, providing for the flow of blanket salt and whose geometry achieves the desired graphite-to-salt volume ratio in the blanket. There were to be 252 of these graphite cylinders, including one in the central location of the reactor that would act as a position for a control rod. The outermost region is the “reflector” consisting of solid graphite cylinders. Some of these cylinders were to be trimmed in order to fit them into the core. This trimming apparently constituted the transition between the prismatic pattern and the metallic reactor vessel.
The manner in which the graphite fuel cells were to be attached to the entrance and exit plena constituted a challenge to the design. The outer graphite structure was to be brazed to a metallic section at the bottom end, and this metallic section would then be welded into the fuel entrance plenum. The inner, “recursive” graphite channel would be fitted to another metallic section using a sliding fit, and then the metallic section would be welded to the exit plenum. Leak-tightness between the exit stream and the entrance stream was not crucial since both streams were of the same composition, differing only in temperature.
The challenge of the design came about in the event that one of the fuel cells needed to be replaced. With the hexagonal geometry of the fuel cells and the very small interstitial space for blanket salt flow between them, there was no room for a long-handled tool to be inserted from the top of the reactor down to the area where the fuel cells connected to the entrance and exit plena. The only option would be to remove the central channel (a simple graphite annular cylinder) and to remove fuel cells, one after another, until the area was reached where the defective fuel cell was present. This would likely involve removing many good fuel cells in order to replace a single defective one. Returning, repairing, and rewelding each of these fuel cells remotely was also a very conceptually challenging task. Each fuel cell would require two welds.
The salt flow patterns in the entrance and exit plena were also never described in detail, but their facile design leads one to consider that they may have embodied many challenges. Since the distance from the large entrance channel to the individual small entrance channels of each fuel cell is different, the pressure drop that could be expected would be highly variable, favoring the fuel cells nearer to the entrance and exit at the expense of those furthest away. It would be desirable to shape the flow through the plena in a way to create uniform pressure drop across each fuel cell, but there is no indication that this was attempted in any of the ORNL two-fluid designs.
The severe challenge of replacing defective fuel cells led ORNL designers to consider instead the replacement of the entire reactor core vessel, with all its graphite internal structures, rather than to have to face the challenge of the individual replacement of defective fuel cells. But this led to great uncertainty. What would be the replacement frequency of defective fuel cells? They had assessed the core replacement frequency due to graphite dimensional change under fast neutron flux and not found it economically excessive, but the uncertainty associated with defective fuel cell replacement was deemed unacceptable. This was the central reason that consideration of one-fluid core designs began.
The importance of removing a defective reactor core influenced the design of the entire reactor module construction. A separate spent reactor vessel cell, labeled “Hot Storage” was included as a place for the spent reactor vessel to cool down after removal and before disassembly, shown in Figure 13. Considerations of disassembly were not included in the final report on the two-fluid design.
Four of the two-fluid reactor designs are compared in Table 1. They show the evolution in core power density from one design to another. One can also observe the changes in transition zone concepts as that issue came into focus. Changes in breeding ratio also accompanied changes in core power density. Later improvements in chemical removal techniques for protactinium would help decouple those effects from one another.
Table 1: Comparison of Two-Fluid Breeder Reactor Designs
- ORNL-3708 salt composition data from Table 1, pg 4.
- ORNL-3936 performance data largely from Table 6.1, pgs 180-181 and Table 6.7, pg 188, also from ORNL-3996, Table 3.1, pgs 37-38 and Table 3.1, pg 41.
- ORNL-4119 performance data largely from Table 9.1, pg 176.
- ORNL-4191 performance data largely from Table 5.2, pg 75.
- Some values for the ORNL-4191 design were taken from the later summary report ORNL-4528.
One-Fluid MSBR Chemical Processing
Research into reductive extraction for bismuth and rare-earths (lanthanides) made ORNL chemical engineers particularly optimistic about the application of these processes to molten-salt breeder reactors. At the same time data was coming back to reactor designers on the dimensional stability of graphite at various temperatures in fast neutrons. The data on dimensional stability was not good, and was calling into question the way that graphite was intended to be used in two-fluid reactor designs.
These pieces of information were being compounded by the pressure to develop a concept for a follow-on breeding experiment to the MSRE, and this required that the eventual MSBR have at least a conceptual design into which the MSBE could trace.
This led ORNL MSRP leadership to undertake a serious shift in the design of the MSBR that we think may have been a mistake. They decided to abandon the two-fluid design, where thorium fertile material and uranium fuel material were kept separate, and to pursue a one-fluid design where thorium and uranium were combined in a single salt. At first blush, one might think that this could lead to a profound simplification of the chemical processing system, since there would no longer be a need to chemically remove U or Pa from the blanket, allow it to decay, and chemically add it to the fuel salt. But the reality of 233Pa’s decay time and its propensity to absorb neutrons, coupled with ORNL’s desire to compete with the LMFBR as a breeder with a short doubling time, led to a more complex chemical processing system than the two-fluid design.
The chemical processing that was proposed for the one-fluid reactor is depicted in Figure 1 and in greater detail in Figure 2. There are several key differences from the chemical processing flowsheet for the two-fluid reactor, and a number of similarities.
In the one-fluid reactor, thorium, uranium, protactinium, and fission products are all mixed together in a single salt. Separation of thorium from lanthanide fission products is rather challenging because of their chemical similarities. (This is the same reason why thorium is generally found in rare-earth deposits.) In each case, protactinium is extracted from the salt so that it can decay outside of the reactor. But in the case of the one-fluid reactor the need to extract protactinium is more pronounced. This is because of the strong desire to reduce the overall reactor inventory of fissile 233U (and shorten the doubling time) the fuel salt (now containing protactinium) is exposed to a greater time-averaged fluence than is the case in the two-fluid design. In the two-fluid design a simple way to reduce the time-averaged fluence is to increase the blanket inventory, but this is not a realistic option in the one-fluid design because of the aforementioned desire to reduce the fissile inventory and doubling time.
The proposed processing scheme was detailed in the report ORNL-TM-3579, Design and Cost Study of a Fluorination-Reductive-Extraction-Metal Transfer Processing Plant for the MSBR, released in May 1972. Fuel salt was first held up for cooling and decay of the shortest lived fission products, then routed to the primary fluorinator, where most of the uranium was removed by fluorination to UF6 using gaseous molecular fluorine (F2) as the fluorination agent. The salt, now stripped of most of its uranium, was routed to an extraction column where metallic bismuth containing lithium and thorium as reductants were contacted with the salt. The remaining uranium, protactinium, and zirconium in the salt were reductively extracted to the bismuth, leaving a salt now that only contained fission products (beyond its base composition of LiF-BeF2-ThF4). This salt entered another reductive extraction column where bismuth containing lithium contacted the salt to remove lanthanide fission products and some thorium. The salt then passed to a reduction column where UF6 was reduced to UF4 in the salt, refueling it and preparing it for return to the reactor. Makeup BeF2 and ThF4 were also added and any residual bismuth was removed from the salt. After a final cleanup step and valence adjustment the purified salt was returned to the reactor.
The bismuth containing some uranium, protactinium, and zirconium was directed to a hydrofluorination column where the metallic solutes in the bismuth were oxidized into their fluoride forms in the presence of a decay salt. The decay salt, containing UF4, PaF4, ThF4, and ZrF4 passed into a decay tank where 233Pa was allowed to decay to 233U. This uranium generated by protactinium decay was removed through fluorination to UF6 and routed to the reduction column to refuel the purified fuel salt.
The bismuth that had been used to carry the proactinium, having been scrubbed of its chemical passengers in the hydrofluorination stage, was routed to the “metal transfer” stage of the processing system where it was combined with bismuth containing lanthanide fission products that had been extracted from the fuel salt. These bismuth streams contacted a salt stream of lithium chloride. Lanthanides transfer to the LiCl but thorium is left behind, accomplishing a decontamination between these two steps. The LiCl is then was successively contacted with streams of bismuth containing metallic lithium reductant which removed the divalent and trivalent lanthanides in separate columns. The bismuth stream containing trivalent lanthanides was hydrofluorinated in the presence of a salt stream that had been designated for waste. The bismuth stream containing divalent lanthanides was combined with the one emerging from the protactinium extraction column and hydrofluorinated into the decay salt. Hence, both the decay salt and the waste salt were contaminated with fission products. Decay salt was the precursor for the waste salt as it was periodically discarded every 220 days. A final fluorination step captured any decayed uranium before discard.
The fluorinators would use F2 as the reagent; the hydrofluorinators would use HF, and the reduction column would use H2. Based on the production and consumption rates, a recycling system for these reagents was proposed. An electrolytic cell would split HF into F2 and H2, which would then be used in the fluorinators and reduction column, respectively. HF emerging from the reduction column would be used in the hydrofluorinators or routed to the electrolytic fluorine cell for production of F2 and H2. Mixed streams of HF and H2 would be separated in an HF distillation system. HF would be sent to the electrolytic cell while H2 would be cleaned up in a caustic scrubber using potassium hydroxide (KOH) in order to capture any residual fluorides. The H2 stream would be recycled to the system but a small amount (5%) would be directed to an alumina absorber, where any fission products like selenium hexafluoride or tellurium hexafluoride would be trapped. The hydrogen stream would also pass through a charcoal absorber to capture noble gases like krypton and xenon before being released up the stack.
Overall, the chemical processing system required for the one-fluid reactor was substantially more challenging than that required for the two-fluid reactor. The fundamental reason for this challenge is the chemical similarity between thorium and the lanthanide fission products, but it was also compounded by the need to extract protactinium rapidly and its connection to fissile inventory.
It should be noted that all of these challenges applied to the goal of a short-doubling-time molten-salt breeder reactor, which was the challenge ORNL faced in the 1960s as these design concepts were being evaluated. If the reactor wasn’t attempting to achieve a high breeding gain, or if it was perhaps not even a breeder at all, but an enriched-uranium-fueled reactor, then many of these challenges might not apply and there would be the potential for tremendous simplification of the chemical processing system. It is important to view these chemical processing systems in the context of the design objectives they were attempting to achieve, which were ambitious then and remain ambitious now.
The capital costs for the chemical processing system of the one-fluid MSBR were given in Table 10 of ORNL-TM-3579 as $35.6M in 1970 dollars. This was based on a 1000-MWe reactor, a reactor fuel volume of 1683 ft3, and a processing cycle time of 10 days. This cost correlates to a cost of $214M in 2013 dollars, or $0.214/watt installed.
Continuous Fluorination Experimental Development
During the time that the one-fluid breeder reactor was the reference design, progress continued to be made in the development of continuous fluorinators, which retained an important position in fuel salt processing. Experimental studies of fluorination of molten salt were carried out in a 1-in.-diam., 72-in.-long nickel fluorinator that allowed countercurrent contact of molten salt with fluorine. In these tests, molten salt (41-24-35 mole% NaF-LiF-ZrF4) containing UF4, was countercurrently contacted with a quantity of fluorine in excess of that required for the conversion of UF4 to UF6. Experiments were carried out with temperatures ranging from 525 to 600°C, UF4 concentrations in the feed salt ranging from 0.12 to 0.35 mole%, and a range of salt and fluorine feed rates. The fraction of the uranium removed from the salt ranged from 97.5% to 99.9%.
Axial dispersion in the salt phase was anticipated to be important in the design of continuous fluorinators, and gas holdup and axial dispersion were measured in columns having diameters ranging from 1 to 6 in. using air and aqueous solutions. Data were obtained for wide ranges of viscosity, surface tension, and superficial gas velocity. Correlations for gas holdup and axial dispersion were developed which were believed to be applicable to countercurrent contact of molten salt and fluorine in a continuous fluorinator. These correlations and the data on uranium removal in the 1-in.-diam continuous fluorinator were used for estimating the performance of larger diameter continuous fluorinators.
The combination of molten salt and fluorine results in a highly corrosive environment, and a future continuous fluorinator will need to protect against corrosion by maintaining a layer of frozen salt on surfaces that would otherwise contact both molten salt and fluorine, preventing molten salt from reaching the surface will allow passivation of the nickel to occur.
The feasibility of maintaining frozen salt layers in gas-salt contactors was demonstrated in tests in a 5-in.-diam, 8-ft-high simulated fluorinator in which molten salt (66-34 mole% LiF-ZrF4) and argon were countercurrently contacted. An internal heat source in the molten region was provided by Calrod heaters contained in a 3/4-in.-diam pipe along the center line of the vessel. A frozen salt layer was maintained in the system with equivalent volumetric heat generation rates of 10 to 55 kW/ft3. For comparison, the heat generation rates in fuel salt immediately after removal from the reactor and after passing through vessels having holdup times of 5 and 30 min are 57, 27, and 12 kW/ft3, respectively.
Operation of a continuous fluorinator with nonradioactive salt required a means for generating heat in the molten salt that was not subject to corrosion. Radio-frequency induction heating in fluorinator simulations was studied using nitric acid as was autoresistance heating using 60-Hz power with molten salt (65-35 mole% LiF-BeF2) in a 6-in.diam fluorinator simulator. Successful operation with auto-resistance heating rates as high as 14.5 kW/ft3 was carried out; the expected power density in processing plant fluorinators is 12 kW/ft3. Autoresistance heating was the preferred method, since it could be used over a wider range of operating conditions and since the electrical power supply is much simpler than that required for induction heating.
Reductive Extraction Experimental Development
Reductive extraction, which was considered as a protactinium removal technique while the two-fluid reactor was the reference design, assumed a much larger role in the one-fluid design. Consequently, ORNL researchers operated a salt-bismuth reductive extraction facility in which uranium and zirconium were extracted from salt by countercurrent contact with bismuth containing reductant. More than 95% of the uranium was extracted from the salt by a 0.82-in.-diam, 24-in.-long packed column. The inlet uranium concentration in the salt was about 25% of the uranium concentration in their one-fluid reference MSBR. These experiments represented the first demonstration of reductive extraction of uranium in a flowing system. Information on the rate of mass transfer of uranium and zirconium was also been obtained in the system using an isotopic dilution method, and HTU values of about 4.5 ft were obtained.
Correlations were developed for flooding and dispersed-phase holdup in packed columns during countercurrent flow of liquids having high densities and a large difference in density, such as salt and bismuth. These correlations, which were verified by studies with molten salt and bismuth, were developed by study of countercurrent flow of mercury and water or high-density organics and water in 1- and 2-in.-diam. columns packed with solid cylinders and Raschig rings varying in size from 1/8 to 1/2 in. Data was also obtained on axial dispersion in the continuous phase during the countercurrent flow of high-density liquids in packed columns, and a simple relation was developed for predicting the effects of axial dispersion on column performance.
The successful operation of salt-metal extraction columns was dependent upon the availability of a bismuth-salt interface detector. To this end, a successful demonstration was made of an eddy-current-type interface detector that consists of a ceramic form on which bifilar primary and secondary coils are wound. Contact of the coils with molten salt or bismuth was prevented by enclosing the element in a molybdenum tube. Passage of a high-frequency alternating current through the primary coil induced a current in the secondary coil whose magnitude was dependent on the conductivities of the adjacent materials; since the conductivities of bismuth and salt are quite different, the induced current reflected the presence or absence of bismuth. The detector appeared to be a practical and sensitive indicator of either salt-bismuth interface location or bismuth level.
Design and development work was initiated on a reductive extraction process facility that would allow operation of the important steps for the reductive extraction process for protactinium isolation. The facility would have allowed countercurrent contact of salt and bismuth streams in a 2-in.-diam., 6-ft-long packed column at flow rates as high as about 25% of those required for processing a 1000-MWe MSBR.
Metal Transfer Process Experimental Development
All aspects of the metal transfer process for the removal of rare earths were demonstrated in an engineering experiment. The equipment consisted of a 6-in.-diam compartmented vessel in which were present about 1 liter each of MSBR fuel carrier salt, bismuth saturated with thorium, and LiCl. The fluoride salt initially contained 147NdF3 at the tracer level and LaF3 at a concentration of 0.04 mole fraction. During the experiment, the rare earths were selectively extracted into the LiCl along with a negligible amount of thorium. Provision was made for circulating the LiCl through a chamber containing bismuth having a lithium concentration of 38 at.%, where the rare earths and thorium were removed. The distribution ratios for the rare earths remained constant during the experiment at about the expected values. About 50% of the neodymium and about 70% of the lanthanum were collected in the Li-Bi solution. The final thorium concentration in the Li-Bi solution was below 5 ppm, making the ratio of rare earths to thorium in the Li-Bi greater than 105 times the initial concentration ratio in the fuel salt and thus demonstrating the selective removal of rare earths from a fluoride salt containing thorium.
A larger metal transfer experiment was put into operation that used salt and bismuth flow rates that are about 1% of the values required for processing a 1000-MWe MSBR, and the preliminary design was carried out for an experiment that would have used a three-stage salt-metal contactor and flow rates that are 5 to 10% of those required for a 1000-MWe MSBR.
Fuel Reconstitution Experimental Development
To reconstitute the fuel salt, UF6 would be directly absorbed in MSBR fuel carrier salt containing UF4, resulting in the formation of soluble non-volatile UF5. Gaseous hydrogen reacts with dissolved UF5 reducing it to UF4
Since both UF6 and UF5 are strong oxidants, experiments were conducted primarily to find a material that was inert to these species. They showed that, at 600°C, nickel, copper, and graphite are not sufficiently inert but that gold is stable both to gaseous UF6 and to salt containing up to 6 wt % UF5. Consequently, subsequent studies were conducted in gold apparatus.
Results from several experiments showed that UF5 dissolved in molten salt slowly disproportionates to UF6 and UF4 and that the rate of disproportionation is second order with respect to the concentration of UF5. The studies also indicate that the solubility of UF6 in the salt is low.
Removal of Bismuth from Fuel Salt
In a processing plant, the fuel salt would be contacted with bismuth containing reductant in order to remove protactinium and the rare earths. It would be necessary that entrained or dissolved bismuth be removed from the salt before it is returned to the reactor, since nickel is quite soluble in bismuth (about 10 wt %) at the reactor operating temperature. Efforts to measure the solubility of bismuth in salt have indicated that the solubility is lower than about 1 ppm, and the expected solubility of bismuth in the salt under the highly reducing conditions that will be used is very low. For these reasons, bismuth can only be present at significant concentrations in the salt as entrained metallic bismuth.
In order to characterize the bismuth concentration likely to be present in the salt after it is contacted with bismuth, ORNL periodically sampled the salt in engineering experiments involving contact of salt and bismuth. The results indicated that the bismuth concentration in the salt in most cases ranged from 10 to 100 ppm after countercurrent contact of the salt and bismuth in a packed-column contactor; however, concentrations below 1 ppm were observed in salt leaving a stirred-interface salt-metal contactor in which the salt and metal phases are not dispersed. One of the difficulties was that of preventing contamination of the samples with small quantities of bismuth during cleaning of the samples and the ensuing chemical analyses.
It was expected that contact of the salt with nickel wool would be effective in removing entrained or dissolved bismuth, since a large nickel surface area can be produced in this manner.
A natural circulation loop constructed of Hastelloy N and filled with fuel salt was operated by the Metals and Ceramics Division for about two years; a molybdenum cup containing bismuth was placed near the bottom of the loop. Reported concentrations of bismuth in salt from the loop (<5 ppm) were essentially the same as those reported for salt from a loop containing no bismuth. No degradation of metallurgical properties for corrosion specimens removed from the loop containing bismuth was noted.
One-Fluid MSBR Core Designs
The MSRP moved from a two-fluid breeder reactor design to a one-fluid design early in 1968 due to the convergence of several developments.
The design studies on the two-fluid reactor had reached a reasonable stopping point in September 1967, and a design power for the MSBE had been selected as 150 MW thermal. In order to improve the performance of the two-fluid reactor, reductive extraction of protactinium from the blanket into bismuth had been investigated. There was hope that protactinium extraction would allow operation of the reactor at a higher core power density while still achieving a high breeding ratio (on the order of 1.08). There had also been investigations into the reductive extraction of lanthanide fission products from the fuel salt into bismuth. The initial results appeared to be promising. A one-fluid breeder would need protactinium to be rapidly and efficiently removed, allowing it to decay to uranium-233 outside of the destructive neutron flux of the reactor. A one-fluid breeder would also have a stronger need than a two-fluid reactor for the rapid removal of lanthanide fission products. Developments in bismuth extraction appeared to provide optimism that such a design change could be accommodated.
One-fluid reactors were also neutronically “leaky” in comparison to two-fluid reactors. There was no absorptive blanket region to productively capture neutrons before they could escape from the reactor vessel. But designers believed they could recreate much of the effect of the blanket in a one-fluid reactor by reducing the moderation (by changing the salt to graphite ratio) and improving the potential of the fertile content of the salt (thorium) to absorb neutrons. By creating two different regions with the positive displacement material (graphite) they could approximate two purposes with a single fluid. It was imperfect, but it seemed to be a positive step.
The drive for the one-fluid design was the challenge of replacing prismatic graphite structures in the design described in ORNL-4191. Due to the complexity of the connections between the graphite fuel cells and the entrance and exit plena, the most acceptable way that they had conceived for graphite replacement was a replacement of the entire core vessel and all the graphite and metallic structures inside it. If the replacement frequency was driven only by the radiation-induced dimensional changes in the graphite this did not appear to be uneconomical. What concerned them was that mechanical failures of individual graphite cells would lead to a replacement frequency for the entire core that was excessive and expensive.
In the one-fluid design, the prismatic graphite structures served only as a moderator and a definition for flow channels; they had no structural function as they did in the two-fluid design. The entrance and exit plena to the core could be simply defined by the placement of graphite prisms. There was no need for a specific plenum structure because there were no fluids to separate. Graphite replacement could be much easier if prisms could be removed individually or in groups without the need to cut them from a plenum or attempt to reconnect them.
None of these realizations was fundamentally incorrect. But there was a degree of over-optimism about the potential performance of the chemical processing system, coupled with an under-appreciation of the loss in versatility, performance, and safety that would be lost with the abandonment of the two-fluid design, that led MSRP designers and engineers to embrace the one-fluid concept.
ORNL-4254, September 1967-February 1968
The first concept for a one-fluid breeder reactor was described in the February 1968 semiannual progress report (ORNL-4254). This design differed in many ways from the previous two-fluid design described in ORNL-4191. The first difference was that it was huge. Rather than a small 250-MWe core design, it was a 2000-MWe (4444 MWt) core that was 18 feet in diameter. Core power density was increased from 20 kW/liter in the ORNL-4191 design back to the 40 kW/liter value that earlier designs had used. Core volume was 4020 ft3, approximately eight times greater than the ORNL-4119 design that had also been designed with a 40 kW/liter average core power density. It was as if designers were attempting to fit eight of their previous reactors in a single vessel in the new design. The core elevation is shown in Figure 1.
The fuel salt, now serving a dual purpose, contained both thorium and uranium tetrafluorides in solution. Its composition was LiF-BeF2-ThF4-UF4 (67.7-20-12-0.3 mole%). The coolant salt continued to be a mixture of sodium fluoroborate and sodium fluoride (92-8 mole%). Graphite was the moderator material and Hastelloy-N the metallic alloy.
With the change in fuel composition came a change in fuel inlet and outlet temperatures. The outlet temperature remained 1300°F, reflecting the material limitation of the Hastelloy-N, but the inlet temperature was raised to 1050°F due to the higher liquidus temperature of the heavier fuel salt. The fuel salt used in the two-fluid design had been a relatively light mixture of primarily LiF-BeF2, with a small amount of UF4 dissolved in it. But the fuel salt of the one-fluid, which included substantial amounts of ThF4 in it, was heavier and required more pumping power.
The reflector was 12 inches in thickness, an increase from previous concepts that could rely on the fertile blanket to productively capture neutrons before they reached the vessel wall. A thick graphite reflector would be a general feature of the one-fluid breeder designs, because of the importance of minimizing neutron loss from the vessel. There were 1760 individual graphite elements in the core.
The flow pattern also changed with the transition from two-fluid to one-fluid. The salt entered at the bottom of the reactor and flowed along the interior and exterior of square graphite prisms with a single channel bored down the center, as shown in Figure 2. Raised “knobs” on the exterior of the prism, running the length of the structure, provided clearance between prisms and created flow channels on the exterior surface of the prism. Another type of prismatic structure was formed from a hollow box inset with a hollow cylinder. This generated a different salt-to-graphite volume ratio and was used in the periphery of the reactor core to create an undermoderated region where neutron absorption would be enhanced. These two different types of graphite structures were needed to create the “core” and “blanket” approximation in a large one-fluid reactor.
The breeding ratio of the new design was uncertain, and was listed as a range of 1.03 to 1.08. Other fuel processing performance aspects of the new design were not given, reflecting the short amount of time the design team had had to make such a significant change.
ORNL-4344, March 1968-August 1968
The August 1968 progress report brough the one-fluid design into greater focus and began to appear more like the “reference” configuration that would ultimately be realized in the next report.
In cross-section, the core region of fuel cells had the shape of an octagon, as shown in Figure 3. An 18-inch-thick graphite reflector lined the interior of the metallic reactor vessel. The concept of a transition region using graphite spheres made a return in this design, filling the gap between the octagonal core and the annular reflector. Average core power density was again reduced to 29 kW/liter and the design thermal power was reduced to 2250 MW for an electrical output of 1000 MWe. Despite the reduction in design power, the reduction in core power density led to larger overall core diameter of 236 inches. The core was squat, without axial graphite blankets to prevent leakage in that direction, as shown in Figure 4.
Protactinium was removed aggressively, on a 3-day cycle, while fission products were removed on a 50-day cycle. Breeding ratio was calculated at 1.077, nearly as good as a two-fluid design.
ORNL-4396, September 1968-February 1969
By the time the semiannual report ending February 1969 was published, the design of the core had very nearly reached its final configuration. As shown in Figure 6, the core prismatic elements now all had a single uniform length, but they were bounded on either end by a thick axial reflector. The radial reflector was also very thick at 30 inches.
The cross-section of the core vessel, shown in Figure 7, showed that the fuel cells retained their overall octagonal design. The radial reflector was divided into wedges that were stacked one on another as blocks to form the reflector structure. The transition from the prismatic core to the annular reflector was achieved by a series of radial graphite ribs or salts that were of variable length. The entire vessel was filled with fuel salt, but flow patterns varied in order to accomplish different goals, such as the keeping of the vessel wall and reflector cooler than the interior of the reactor.
The core power density was reduced to 22.2 kW/liter and the reactor vessel was 270 inches in diameter, with 1412 graphite elements therein set on a 4-inch pitch.
Fuel was processed for protactinium on a 3-day cycle and fission products on a 30-day cycle. Breeding ratio was anticipated to be 1.06.
ORNL-4541, MSBR Conceptual Design Study, June 1971
The “reference” one-fluid reactor design was written up in a separate report, ORNL-4541, published in June 1971. The reactor design described therein was nearly identical to the one revealed in ORNL-4396. One of the only changes was a modification to the fuel processing schedule, with ORNL-4541 projecting a protactinium and fission-product removal time of 10 days.
ORNL-4548, September 1969-February 1970
Later an attempt to simplify the entire core design was briefly considered in ORNL-4548. As shown in Figure 9, the core vessel would be made spherical and lined with thick graphite wedges to act as the reflector. The interior volume would be filled with graphite spheres and fuel salt. The advantage of this design was that the addition and removal of graphite spheres was considered simple. The fatal disadvantage of the design, however, was that the necessary volume ratio between the fuel salt and the graphite could not be achieved with loose graphite spheres, and so the idea was not pursued further.
One-Fluid MSBR Core Design Summary
Four of the one-fluid reactor designs are compared in Table 1. They show that most design aspects remained fairly stable in the evolution of one-fluid designs, with the exception of the steady reduction in core power density. Notable is the steady increase in graphite reflector thickness that accompanied between understanding of the neutronics of one-fluid breeder designs.
Table 1: Comparison of One-Fluid Breeder Reactor Designs
- ORNL-4254 performance data largely from Table 5.2, pg 59.
- ORNL-4344 performance data largely from Table 5.1, pg 56.
- ORNL-4396 performance data largely from Table 5.1, pg 54.
- ORNL-4541 performance data largely from Table S.1, pgs xi-xv.