This is a very nice summary of how ORNL viewed the origins, evolution, and progress of their fluoride reactor research in 1972. It comprises chapter 2 of the ORNL report 4812: The Development Status of Molten-Salt Breeder Reactors.
2. EVOLUTION AND DEVELOPMENT OF MOLTEN-SALT REACTORS
P. N. Haubenreich
When the idea of the breeder was first suggested in 1943, the rapid and efficient recycle of the partially spent core was regarded as the main problem . This problem, which is still crucial in breeder economics, was attacked in two ways—by striving for very long burnup and by seeking to simplify the entire recycle operation. The latter pursuit inevitably led to consideration of fluid-fueled reactors as the ultimate in fuel cycle simplification.
Neutron-multiplying systems consisting of aqueous solutions and slurries were investigated soon after the discovery of nuclear fission, and the first effort toward a fluid-fueled breeder was based on these systems, in which the fluid is both the fuel and the moderator [2, pp. 1-9]. The Homogeneous Reactor Program, organized at ORNL in 1949, had as its objective a reactor with a uranyl sulfate-D2O solution core and a thorium oxide-D2O slurry blanket, separated by a Zircaloy core tank. This concept, with its superior neutron economy, offered good 233U breeding performance.
About 1950 the idea of a very different fluid-fuel reactor for power generation emerged at Brookhaven National Laboratory from studies on low-melting alloys and slurries of uranium and thorium in liquid metal. This was the liquid-metal-fuel reactor concept [2, pp. 699-929]. A version of the LMFR using graphite moderator, U-Bi solution core, and ThO2-Bi slurry blanket appeared capable of breeding.
Meanwhile yet another fluid-fueled reactor had been conceived for an altogether different purpose—aircraft propulsion. Several different concepts of compact reactors were being considered for generating heat to be used in a jet engine. The Oak Ridge idea was to use a high-temperature liquid fuel that could be circulated to remove heat from the core and be drained for refueling. Experiments to investigate molten-salt fuels were begun in 1947, and 3 years later molten fluorides were chosen for the main effort of the Aircraft Nuclear Propulsion (ANP) program at ORNL . The fluorides were particularly well suited because they offer low vapor pressure at jet-engine temperatures, reasonably good heat transfer properties, and immunity to radiation damage, and they do not react violently with air or water. (See Chapter 5.) A small reactor, the Aircraft Reactor Experiment, was built that used a fuel mixture of NaF, ZrF4, and UF4 circulating in Inconel tubing through a moderator assembly of BeO blocks [2, pp. 673-80]. In 1954 the ARE was operated successfully for 9 days at outlet temperatures ranging to above 1600ºF (1150 K) and powers to 2.5 MW(t) in investigations of the nuclear dynamics of the circulating fuel system.
It was recognized from the outset that molten-salt reactors might be attractive for civilian power applications, and in 1956 a group was formed at ORNL to study the characteristics, performance, and economics of molten-salt reactors for central station power generation . A wide variety of configurations were considered, and some were found that promised low power costs while breeding in the thorium-233U cycle. Thus when the effort to develop a molten-salt aircraft reactor was stopped in 1957, the molten-salt reactor concept survived as a potential civilian power reactor.
Relation to Other Fluid-Fuel Programs
Early in 1959 a task force assembled by the AEC made a comparative evaluation of the three fluid-fuel reactor concepts then being pursued. The conclusion was that the molten-salt reactor, although limited in potential breeding gain, had “the highest probability of achieving technical feasibility” . Soon thereafter work on the aqueous homogeneous and liquid-metal-fuel reactors was discontinued, leaving the molten-salt reactor as the lone fluid-fuel breeder concept still being supported by the USAEC.*
Although the Molten-Salt Reactor Program, as such, was relatively young, there was an extensive technological base from the ANP program, where $60 million had been invested in molten-salt reactor technology. Some of this had gone for developments specific to the compact aircraft configuration, but a large fraction of the technology was equally applicable to the civilian power reactors that were being envisioned. The physical chemistry of interesting fluoride salt mixtures had been explored, and a container alloy had been developed that was especially compatible with fluoride salt mixtures and which had significantly higher strength than Inconel at the 1500-1600ºF temperatures required in an aircraft reactor. Originally called INOR-8, this alloy is now generally known as Hastelloy-N. Techniques for producing, purifying, and analyzing fluoride mixtures had been worked out, and considerable experience was gained in handling the ARE fuel. The fluoride volatility process was developed and was successfully used to recover the uranium from the ARE fuel in 1957-58.
In addition to the generally applicable ANP work, there was some spinoff to the molten-salt technology from the aqueous homogeneous reactor and liquid-metal-fuel reactor programs. The Homogeneous Reactor Program had built and operated two reactors using circulating aqueous fuel solutions at 250-300ºC. Considerable maintenance was required on radioactive parts of these reactors, and one significant contribution to reactor technology was this experience with maintenance of highly radioactive systems . A chemical processing scheme explored for the liquid metal fuel reactor involved molten salts and molten bismuth. The experience of this effort, and the general background of experience with molten bismuth, proved valuable when extraction systems involving molten bismuth became the heart of the processing concept for MSBR’s.
Early MSBR Concepts
In the early days of the Molten-Salt Reactor Program, serious consideration was given to homogeneous reactors in which the core contained nothing but salt. These ideas were abandoned after calculations showed that the limited moderation by likely fluoride salt constituents alone would result in a thermal reactor with inferior breeding performance. Breeding appeared possible in intermediate-spectrum reactors, but their gains were not high enough to compensate for their higher fissile inventories. Studies of fast-spectrum molten-salt reactors (using chloride salts) indicated good breeding ratios, but fissile inventories were excessive unless unconven-tional heat transfer systems were used to minimize holdup outside of the core.
After experiments showed that bare graphite could probably be used in the core of a molten-salt reactor, MSRP efforts concentrated on graphite-moderated reactors having well-thermalized neutron spectra and low fissile inventories. Two general types were considered—single-fluid reactors in which thorium and uranium were combined in one salt, and two-fluid reactors in which UF4-bearing fuel salt was separated from fertile salt containing ThF4. In any case the diluent fluoride mixture would be 7LiF-BeF2 instead of the NaF-ZrF4 mixture used in the aircraft reactors; the 7LiF-BeF2 absorbed fewer neutrons and dissolved more ThF4 without excessive liquidus temperatures. The single-fluid reactor was relatively simple and promised low power costs, but breeding appeared to be impractical because of neutron leakage and losses to protactinium and fission products . (At that time it was not clear that Pa and fission products could be separated economically on a very short cycle.) The two-fluid reactor could be designed with a fertile blanket to reduce leakage, and Pa losses would be reduced because the fertile salt would be at a lower average flux. The only processing required for the fertile salt was fluorination to recover the bred uranium. The fuel salt could be processed by a combination of fluorination and an aqueous process. The two-fluid reactor was more complex in that it used two salts that had to be kept separate, but it did offer attractive breeding performance.
By 1960 a fairly clear picture of a family of molten-salt reactors had emerged. The technical feasibility appeared to be on a sound footing—a compatible combination of salt, graphite, and container material—but a reactor was needed to really prove the technology. That was the purpose of the Molten-Salt Reactor Experiment: to demonstrate that some of the key features of the proposed molten-salt power reactors could be embodied in a practical reactor that could be operated safely and reliably and be maintained without excessive difficulty. For simplicity it was to be a fairly small, one-fluid reactor operating at 10 MW(t) or less, with heat rejection to the air via a secondary salt.
The MSRE flowsheet is shown as Figure 2.1. Figure 2.2 shows some detail of the 5-ft-diameter reactor vessel. The fuel was LiF-BeF2-ZrF4-UF4 (64-30-5-1 mole %), the secondary salt was LiF-BeF2 (66-34 mole %), the moderator was grade CGB graphite, and all other parts contacting salt were of Hastelloy-N. The bowl of the fuel pump was the surge space for the circulating loop, and here about 50 gpm of fuel was sprayed into the gas space to allow xenon and krypton to escape from the salt. Also in the pump bowl was a port through which salt samples could be taken or capsules of concentrated fuel enriching salt (UF4-LiF or PuF3) could be introduced. The fuel system was located in sealed cells, laid out for maintenance with long-handled tools through openings in the top shielding. A tank of LiF-BeF2 salt was used to flush the fuel circulating system before and after maintenance. In a cell adjacent to the reactor was a simple facility for bubbling gas through the fuel or flush salt: H2-HF to remove oxide, F2 to remove uranium as UF6. References 9, 10, and 11 provide more detailed descriptions of the reactor and processing plant.
Development and Construction
Most of the MSRP effort from 1960 through 1964 was devoted to design, development, and construction of the MSRE. Production and further testing of graphite and Hastelloy-N, both in-pile and out, were major development activities. Others included work on reactor chemistry, development of fabrication techniques for Hastelloy-N, development of reactor components, and remote-maintenance planning and preparations. (A convenient summary of developments through the end of major construction is given in reference 12.)
Before the MSRE development began, tests had shown that salt would not permeate graphite in which the pores were very small. Graphite with the desired pore structure was available only in small, experimentally prepared pieces, however, and when a manufacturer set out to produce a new grade (CGB) to meet the MSRE requirements, difficulties were encountered [12, pp. 373-309]. A series of pitch impregnations and heat treatments produced the desired high density and small pore structure, but in the final steps occasional cracks appeared in many of the 2 ¼-in. square bars. Apparently the cracks resulted because the structure was so tight that gases from the pyrolysis of the impregnant could not escape rapidly enough. Tests showed, however, that the cracks did not propagate, even when filled with salt and subjected to repeated freeze-thaw cycles. After analysis showed that heating in salt-filled cracks would not be excessive, the graphite was accepted and used in the MSRE.
The choice of Hastelloy-N for the MSRE was on the bases of the promising results of tests at ANP conditions and the availability of much of the required metallurgical data.* Development for the MSRE generated the further data required for ASME code approval. It also included preparation of standards for Hastelloy-N procurement and for component fabrication. Material for the MSRE, amounting to almost 200,000 lb in a variety of shapes, was produced commercially. After weld-cracking in experimental heats was overcome by minor composition changes, there was no difficulty in obtaining acceptable material. Requests for bids on component fabrication went to several companies in the nuclear fabrication industry, but all declined to submit lump-sum bids because of lack of experience with the new alloy. Consequently all major components were fabricated in AEC-owned shops at Oak Ridge and Paducah [12, pp. 63-52]. After appropriate procedures were worked out, Hastelloy-N fabrication presented no unusual problems.
At the time that design stresses were set for the MSRE, the few data that were available indicated that the strength and creep rate of Hastelloy-N were hardly affected by irradiation. An arbitrary allowance was made for possible effects, however, by establishing design stresses 20% below Code values for unirradiated Hastelloy-N. After the construction was well along, the stress-rupture life and fracture strain were found to be drastically reduced by thermal-neutron irradiation. The MSRE stresses were reanalyzed, and it was concluded that the reactor would have adequate life to reach its goals. At the same time a program was launched to improve the resistance of Hastelloy-N to the embrittlement. (See Chapter 7 and reference 13.)
An extensive out-of-pile corrosion test program was carried out for Hastelloy-N [12, pp. 334-343] which indicated extremely low corrosion rates at MSRE conditions. Capsules exposed in the Materials Testing Reactor showed that salt fission power densities of more than 200 W/cm3 had no adverse effects on compatibility of fuel salt, Hastelloy-N, and graphite. Fluorine gas was found to be produced by radiolysis of frozen salts, but only at temperatures below about 100ºC [12, pp. 252-257]. The results of this program are described in some detail in Chapter 5.
Components that were developed especially for the MSRE included flanges for 5-inch lines carrying molten salt, freeze valves (an air-cooled section where salt could be frozen and thawed), flexible control rods to operate in thimbles at 1200°F, and the fuel sampler-enricher [12, pp. 167-190]. Centrifugal pumps were developed similar to those used successfully in the aircraft reactor program, but with provisions for remote maintenance, and including a spray system for xenon removal. Remote maintenance considerations pervaded the MSRE design, and developments included devices for remotely cutting and brazing together 1½-inch pipe, removable heater-insulation units, and equipment for removing specimens of metal and graphite from the core.
The MSRE development program did not include reactor physics experiments or heat transfer measurements. There was enough latitude in the MSRE that deviations from predictions would not compromise safety or accomplishment of the objectives of the MSRE.
Construction of the primary system components and alterations of the old ARE building (which had been partly remodeled for a proposed 60-MW(t) aircraft reactor) were started in 1962. Installation of the salt systems was completed in mid-1964. ORNL was responsible for quality assurance, planning, and management of construction . The primary systems were installed by ORNL forces; subcontractors modified the building and installed ancillary systems.
Operation of the MSRE spanned 5 years, from the loading of salt in 1964 through the end of nuclear operation in December, 1969. As described in references 9 and 15, all of the objectives of the experiment were achieved during this period.
Checkout and prenuclear tests included 1000 hr of circulation of flush salt and fuel carrier salt. Nuclear testing of the MSRE began in June 1965, with the addition of enriched 235U as UF4-LiF eutectic to the carrier salt to make the reactor critical. After zero-power experiments to measure rod worth and reactivity coefficients , the reactor was shut down and final preparations made for power operation. Power ascension was delayed when vapors from oil that had leaked into the fuel pump were polymerized by the radioactive offgas and plugged gas filters and valves. Maximum power, which was limited to 7.4 MW(t) by the capability of the heat-rejection system, was reached in May 1966.
After two months of high-power operation, the reactor was down for 3 months because of the failure of one of the main cooling blowers. Some further delays were encountered because of offgas line plugging, but by the end of 1966 most of the startup problems were behind. During the next 15 months, the reactor was critical 80% of the time, with runs of 1, 3, and 6 months that were uninterrupted by a fuel drain. By March, 1968, the original objectives of the MSRE had been accomplished, and nuclear operation with 235U was concluded.
By this time, ample 233U had become available, so the MSRE program was extended to include substitution of 233U for the uranium in the fuel salt and, operation to observe the new nuclear characteristics. Using the on-site processing equipment, the flush salt and fuel salt were fluorinated to recover the uranium in them as UF6 . 233UF4-LiF eutectic was then added to the carrier salt, and in October 1968, the MSRE became the world's first reactor to operate on 233U.
The 233U zero-power experiments and dynamics tests confirmed the predicted neutronic characteristics.* An unexpected consequence of processing the salt was that its physical properties were altered slightly so that more than the usual amount of gas was entrained from the fuel pump into the circulating loop. The circulating gas and the power fluctuations that accompanied it were eliminated by operating the fuel pump at slightly lower speed. Operation at high power for several months permitted very accurate measurement of the capture-to-fission ratio, for 233U in this reactor, completing the objectives of the 233U operation.
In the concluding months of operation, xenon stripping, deposition of fission products, and tritium behavior were investigated. The feasibility of using plutonium in molten-salt reactors was emphasized by adding PuF3 as makeup fuel during this period.
After the final shutdown in December 1969, the reactor was left in standby for almost a year. Then a limited examination program was carried out, including a moderator bar from the core, a control rod thimble, heat exchanger tubes, parts from the fuel pump bowl, and a freeze valve that had developed a leak during the final shutdown. The radioactive systems were then closed to await ultimate disposal.
The broadest and perhaps most important conclusion from the MSRE experience is that this was quite a practical reactor. It ran for long periods of time, yielding valuable information, and when maintenance was required it was accomplished safely and without excessive delay.† The remarkable performance of the MSRE clearly shows that with proper design and careful construction and operation, the unusual features of an MSR in no way compromise its safety and dependability.
In many regards, the MSRE served to confirm expectations and predictions . For example, we had confidently expected the observed immunity of the fuel salt to radiation damage, the complete absence of attack on the graphite, and the very minor general corrosion of the Hastelloy-N. Noble gases were stripped from the fuel salt by the simple spray system even better than anticipated, reducing the 135Xe poisoning by a factor of about 6. The bulk of the fission product elements remained stable in the salt. Additions of uranium and plutonium to the salt during operation were quick and uneventful, and the recovery of uranium by fluorination was quite efficient. The neutronics, including critical loading, reactivity coefficients, dynamics, and long-term reactivity changes, agreed very closely with prior calculations.
In other areas, the operation resulted in improved data or helped to reduce uncertainties. The 233U capture-to-fission ratio in a typical MSR neutron spectrum is an example of basic data that were improved. The effect of fissioning on the redox potential of the fuel salt was resolved. The deposition of some elements (“noble metals”) was expected, but the MSRE provided quantitative data on relative deposition on graphite, metal, and liquid-gas interfaces. Heat transfer coefficients measured in the MSRE agreed very closely with conventional design calculations (using correct values for salt properties) and did not change over the life of the reactor. Limitation of oxygen access to the salt proved quite effective, and the tendency of fission products to be dispersed from contaminated equipment during maintenance was less than we had anticipated.
Operation of the MSRE provided some insights into the unusual problem of tritium in a molten-salt reactor. It was observed that about 6-10% of the calculated 54 Ci/day production diffused out of the fuel system into the containment cell atmosphere and another 6-10% reached the air through the heat removal system . The fact that these fractions were not higher indicated that something (probably oxide coatings) partially negated the easy transfer of tritium through hot metals.
The one quite unexpected finding of great importance was the shallow inter-granular cracking observed in all metal surfaces exposed to the fuel salt. This was first noted in the specimens that were removed from the core at intervals during the reactor operation. Post-operation examination of pieces of a control-rod thimble, heat-exchanger tubes, and pump bowl parts revealed the ubiquity of the cracking and emphasized its importance to the MSR concept. Further investigations and possible consequences are discussed in Chapters 7 and 14 of this report.
Recent Molten-Salt Reactor Concepts
Since the MSRE, the Molten-Salt Reactor Program has been a technology program, not focused on building a particular reactor, but seeking to identify and accomplish the developments that are needed before molten-salt breeder reactors can become a reality . In the furtherance of this program, efforts on conceptual design have been essential in defining the needs for development, while experimental findings, in turn; shape the concept. This section describes the reactor concepts that have been considered in the course of this intertwined process.
As described previously, at the time that the MSRE was conceived, the two-fluid reactor, despite its relative complexity, seemed to hold the most promise as a breeder. During the early years of the MSRE, relatively little effort was devoted to refinement of conceptual designs. Basic chemistry studies continued, however, and led in 1964 to an important development that simplified the processing in the two-fluid breeder plant [12, p. 309]. This was the separation of rare-earth fluorides from LiF and BeF2 by distillation at 1000ºC. (The practicality was later demonstrated with a portion of the MSRE fuel .) Thus it was, when the MSRE settled into operation, that design efforts focused on the two-fluid concept.
Studies at first indicated outstanding resource utilization, mainly because of an extremely low specific inventory of about 0.8 kg fissile/MW(e) . Then in 1967, when irradiation of graphite to very high neutron fluences revealed more rapid dimensional changes than had been projected, the two-fluid concept was dealt a severe blow. Accommodation of the differential growth of the graphite made the core design and assembly so complex that it seemed necessary to replace the entire reactor vessel and its contents whenever the graphite core tubing became unserviceable. The reference design for a 1000-MW(e) plant included four 556-MW(t) reactors that could be replaced at staggered intervals to improve plant availability . The two-fluid reactor could be scaled down without seriously affecting breeding gain, but performance was hurt, because in order to extend graphite life to 8 full-power years, the power density in the reference design was reduced from 80 W/cm3 to 20 W/cm3, at the expense of raising the specific inventory to 1.3 kg fissile/MW(e).
At about this time a chemical processing development occurred that greatly improved the prospect for economical breeding in a simpler, single-fluid reactor. This was the laboratory demonstration of the basic chemical steps in a continuous process for removing protactinium and uranium from molten fluoride mixtures that contain thorium fluoride. When LiF-BeF2-ThF4-UF4-PaF4 salt was contacted with molten bismuth containing dissolved thorium and lithium, first the U and then the Pa were reduced and passed from the salt into the liquid metal. The Pa could be sequestered until it decayed, while the uranium was returned to the fuel salt by electrolysis. This reductive-extraction process, which could draw upon technology developed at Argonne and Brookhaven for processing fast reactor fuels and the U-Bi fuel of an LMFR, appeared practicable for continuous separation of protactinium. It also appeared that it might be adaptable to removal of rare-earth fission products, thus permitting a relatively simple processing plant that would keep breeding losses due to Pa and fission products to acceptably low levels.
After the recognition of the Pa-removal possibility, one-fluid breeder core designs were explored more thoroughly than in earlier surveys. As a result, it was found that breeding performance could be significantly improved by a scheme proposed several years before. By decreasing the graphite fraction in the outer part of the core, the neutron spectrum there can be hardened, increasing the fraction of captures in thorium, while the fission neutron production is more concentrated in the inner, well-moderated part of the core. The effect is to reduce neutron leakage, which had always been a significant factor in one-fluid breeders.
The combined effect of the new Pa-removal system and the improved core design was to increase the breeding ratio that could be achieved economically in a one-fluid breeder to about 1.05-1.07. Consequently, the resource utilization characteristics became acceptable, and in 1968 the major emphasis of the MSRP was shifted to the development of the simpler single-fluid breeder reactor. Later, substantial improvements were made in the processing system, including elimination of the electrolysis cells, storage of Pa in salt instead of in bismuth, and development of a more efficient process for removal of rare earths. (See Chapter 11.)
Throughout this evolution, the primary long-range objective of the MSRP remained the same—efficient breeding in the thorium-233U cycle. This pursuit led by 1970 to the reference conceptual design of a 1000-MW(e) MSBR plant  described in Chapter 3, which is the focus of the MSRP development effort. In addition, ORNL has recently investigated several other versions of one-fluid molten-salt reactors that also meet the essential criteria of safety, reliability, and low power costs, while offering one advantage or another relative to the reference MSBR.
One alternate offers a way around the necessity of replacing the core at 4-year intervals. This difficult job is avoided by simply making the core of the breeder large enough and the damage flux low enough that the core graphite will last the 30-year life of the plant. Such a "permanent-core" reactor can have a breeding ratio as high as in the reference design, but the large core means a greater inventory and longer doubling time .
Another possible simplification is the elimination of most of the chemical processing. If it should happen in a breeder plant that protactinium and fission-product removal were stopped, the reactor could continue to operate for months, or even years, as a near-breeder with only the addition of enriched uranium. Alternatively, the reactor could be built as a converter with no processing (other than perhaps oxide removal). Some studies indicate that an economical mode of operation would be to run for 6 equivalent full-power years, recover the uranium by batch fluorination (as demonstrated in the MSRE), discard the salt with the fission products, and resume with fresh carrier salt. Either 235U or plutonium could be used for startup and feed of such a molten-salt converter reactor.
A limited amount of conceptual design was done on a 350-MW(e) converter that positively overcomes the problems of graphite replacement and tritium containment with a minimum of additional development. The core is large enough so that presently available graphite would last 30 years, and tritium containment is ensured by using an intermediate salt loop containing Hitec . The Hitec, a KNO3-NaNO¬3-NaNO2 mixture, reacts with tritium to form water, which could be stripped continuously from the salt. Because the nitrate-nitrite mixture would precipitate uranium and might react violently with graphite if it leaked into the fuel system, a compact loop containing LiF-BeF2 (which is quite compatible with the fuel) is interposed between the fuel and Hitec systems. The disadvantages of the extra loop and temperature limitations on Hitec are weighed against simplification of the steam system because of the relatively low melting point of Hitec (288°F). In any event this study is an offshoot from the main line of ORNL effort, which is directed at the high-performance breeder.
In considering the most expeditious route to the ultimate molten-salt breeder, we have devoted some attention to the conceptual design of what we call the Molten-Salt Breeder Experiment (MSBE). This reactor would be designed to operate under conditions at least as severe as those in a large plant. Design studies show that a graphite damage flux twice that in the reference MSBR and power density, salt composition, and protactinium and fission-product concentrations like those in the reference design could be attained in a 150-MW(t) reactor with a 4-ft core in a 7.5-ft vessel . The MSBE would produce steam and would include a complete processing facility. Although its small dimensions limit its breeding ratio to about 0.96, the MSBE would provide a definitive test of all the basic equipment and processes required for a breeder.
Most of the work on molten-salt reactors is now, as in the past, concentrated in the Molten-Salt Reactor Program at ORNL, but some significant activities are going on elsewhere. This section outlines the MSRP activities (which are fully described in other chapters) and briefly describes the other programs.
USAEC Molten-Salt Reactor Program
The Molten-Salt Reactor Program is included with the LMFBR and the gas-cooled fast breeder in the AEC's “high-gain breeder” development program. A total of $64 million has been spent by the MSRP from its inception in 1957 through fiscal year 1972, including the $10 million cost of designing and building the MSRE. The MSRP also has benefited from basic work carried out under the AEC's physical research programs. The MSRP budget for FY-1973 is $5 million, now allocated approximately as follows:
12% Reactor design and analysis
18% Reactor engineering technology
20% Reactor metals
16% Chemistry and analytical chemistry
22% Fuel processing
4% Processing materials
The MSRP design and analysis effort includes evaluation of the potential of molten-salt reactors and preparation of reference designs and alternatives that define the research and development needs. This has been cut back in recent years, and nearly half of the current effort is a study being done under subcontract by an industrial group headed by Ebasco Services.
The bulk of the engineering development effort is presently devoted to two MSRE-scale loops: the Coolant Salt Technology Facility and the Gas Systems Test Facility, described in Chapter 8. Also under this heading is a study of molten-salt steam generators being carried out by Foster-Wheeler under subcontract to ORNL.
Reactor chemistry is mainly concerned at present with the behavior of tritium and certain fission products. Analytical development is aimed at the techniques needed for the operation of a breeder and its processing plant.
Reactor metals development is focused on improving Hastelloy-N (or finding an alternative) to overcome the problems of neutron embrittlement and surface cracking by fission products. Sealing to exclude xenon and better structure for longer life are being investigated in the graphite program. Processing materials work is now centered on molybdenum fabrication, with some effort on graphite.
The emphasis in fuel processing is on reductive extraction—the chemistry of fluoride-bismuth and chloride-bismuth systems and engineering equipment to exploit this chemistry. Fluorination and fuel reconstitution are being developed, and alternate processes such as oxide precipitation are being studied.
There have been two privately funded conceptual design studies and evaluations of MSR's. The first was by the Molten-Salt Breeder Reactor Associates (MSBRA), headed by the engineering firm of Black & Veatch and including five midwestern utilities. The MSBRA study identified problem areas but concluded that the economics of molten-salt reactors were attractive relative to light-water reactors, and favored a program leading to early commercial application of a molten-salt converter . Since the conclusion of their study in 1970 the MSBRA has been relatively inactive.
The second privately funded organization is the Molten-Salt Group (MSG), whose formation was announced in 1969 [27, p. 1-1]. The MSG is headed by Ebasco Services and includes 5 other industrial firms* and 15 utility companies . In the fall of 1971 the MSG completed an evaluation of the state of MSR technology  and a critique of the ORNL 1000-MW(e) breeder design . They concluded in the first report that the existing technology was sufficient to justify construction of a demonstration plant that would breed, provided the processing works as intended, but that its maintainability, reliability, costs, and plant life could not be predicted reliably from the existing technology [28, p. 6]. In the second report, they concluded that the ORNL reference design embraced some technological difficulties, but was a suitable departure point for exploring MSBR technology [29, p. 4]. The MSG members have agreed to support studies of a demonstration plant and alternate molten salt reactor concepts and an updating of their MSBR technology evaluation. These studies are proceeding concurrently with the AEC-funded MSBR design study that is being done by part of the Group's working force.
There have been other indications of interest in molten-salt reactors by industry and utilities. Many have sent representatives to MSRP annual information meetings, and several have made private studies of their possible role in MSR development. A few have assigned staff members to work in the MSRP (for up to 2 years), and materials producers have contributed by providing experimental graphite and alloys for evaluation.
The Indian Department of Atomic Energy is interested in molten-salt reactors in connection with their vast thorium resources and the anticipated availability of plutonium in India for MSR startup. A small program of research on the MSBR concept has been underway at the Bhabha Atomic Research Centre since 1969. The program is mostly concerned with investigations of the chemistry of melts containing PuF3. The USAEC and the DAE have been exchanging molten-salt information, and an agreement providing for somewhat broader cooperation is being considered.
Euratom has supported research on various topics pertinent to molten-salt reactor technology. From 1964 to 1966 there was an exchange of molten-salt information through a formal agreement with the USAEC, but the exchange is now relatively inactive. Development work involving construction and operation of a molten-salt steam generator that was begun with Euratom support is continuing at Delft University, however. Other European molten-salt reactor work was conducted at Kernforschungsanlage Julich in 1963 to 1967, mostly in connection with the epithermal reactor concept, MOSEL . This has been discontinued.
The UKAEA has supported a small effort on molten-salt reactors for several years. Studies and a limited amount of experimental work are continuing, with emphasis on chloride-fueled, lead-cooled fast reactors.
References for Chapter 2
1. A.M. Weinberg, “Molten-Salt Reactors,” Nuclear Applications and Technology. 8, 105 (1970).
2. J.A. Lane, H.G. MacPherson, and F. Maslan, Fluid Fuel Reactors, Addison-Wesley, Reading, Mass., 1955.
3. H.G. MacPherson, “Molten-Salt Reactors,” Proceedings of the International Conference on Constructive Uses of Atomic Energy, Washington, Nov. 1968, American Nuclear Society, 1969, p. 111.
4. M.W. Rosenthal, P.R. Kasten, and R.B. Briggs, “Molten-Salt Reactors—History, Status and Potential,” Nuclear Applications and Technology. 8, 107 (1970).
5. Report of the Fluid Fuel Reactors Task Force, TID-8507 (February 1959).
6. J.J. Went and M.E.A. Hermans, “The KEMA Suspension Test Reactor,” Fourth UN International Conference on the Peaceful Uses of Atomic Energy, Geneva, A/CONF 49/P/020 (1971).
7. P.N. Haubenreich, “Two Years of HRE-2 Operation,” Nuclear Science and Engineering. 5, 467 (1960).
8. L.G. Alexander, W.L. Carter, C.W. Craven, D.B. Janney, T.W. Kerlin, and R. Van Winkle, Molten-Salt Converter Reactor, ORNL-TM-1060 (September 1965).
9. P.N. Haubenreich and J.R. Engel, “Experience with the MSRE,” Nuclear Applications and Technology. 8, 118 (1970).
10. R.C. Robertson, MSRE Design and Operations Report, Part I, Description of Reactor Design, ORNL-TM-728 (January 1965).
11. R.B. Lindauer, Processing of the MSRE Flush and Fuel Salts, ORNL-TM-2578 (August 1969).
12. MSR Program Semiannual Progress Report, July 31, 1964, ORNL-3708.
13. H.E. McCoy et al., “New Developments in Materials for Molten-Salt Reactors,” Nuclear Applications and Technology. 8, 156 (1970).
14. B.H. Webster, Quality-Assurance Practices in Construction and Maintenance of the MSRE, ORNL-TM-2999 (April 1970).
15. M.W. Rosenthal, P.N. Haubenreich, H.E. McCoy, and L.E. McNeese, “Current Progress in Molten-Salt Reactor Development,” Atomic Energy Review IX, 601-50 (1971).
16. B.E. Prince, S.J. Ball, J.R. Engel, P.N. Haubenreich, and T.W. Kerlin, Zero-Power Physics Experiments on the MSRE, ORNL-4233 (February 1968).
17. R.B. Briggs, “Tritium in Molten-Salt Reactors,” Reactor Technology, 14, 335-42 (Winter 1971-72).
18. H.G. MacPherson, “Development of Materials and Systems for the Molten-Salt Reactor Concept,” Reactor Technology. 15, 136-155 (Summer 1972).
19. J.R. Hightower, L.E. McNeese, B.A. Hannaford, and H.D. Cochran, Low-Pressure Distillation of a Portion of the Fuel Carrier Salt from the MSRE, ORNL-4577 (August 1971).
20. P.R. Kasten, E.S. Bettis, and R.C. Robertson, Design Studies of 1000-MW(e) Molten-Salt Breeder Reactors, ORNL-3996 (1966).
21. R.C. Robertson, R.B. Briggs, O.L. Smith, and E.S. Bettis, Two-fluid Molten-Salt Breeder Reactor Design Study (Status as of Jan. 1, 1968), ORNL-4528 (1970).
22. Conceptual Design Study of a Single-Fluid Molten-Salt Breeder Reactor, ORNL-4541 (1971).
23. MSR Program Semiannual Progress Report, August 31, 1970, ORNL-4622.
24. E.S. Bettis, L.G. Alexander, and H.L. Watts, Design Studies of a Molten-Salt Reactor Demonstration Plant, ORNL-TM-3832 (June 1972).
25. J.R. McWherter, Molten-Salt Breeder Experiment Design Bases, ORNL-TM-3177 (November 1970).
26. Molten-Salt Breeder Reactor Associates Staff, Final Report, Phase I Study—Project for Investigation of Molten-Salt Breeder Reactor, Black & Veatch Consulting Engineers, Kansas City, Mo. (1970).
27. 1000-MW(e) Molten-Salt Breeder Reactor Conceptual Design Study, Final Report—Task I, Ebasco Services, Inc., New York, February 1972.
28. Molten-Salt Reactor Technology, Technical Report of the Molten-Salt Group, Part I, Ebasco Services, Inc., December 1971.
29. Evaluation of a 1000-MWe Molten-Salt Breeder Reactor, Technical Report of the Molten-Salt Group, Part II, Ebasco Services, Inc., October 1971.
30. P.R. Kasten, “The MOSEL Reactor Concept,” Third UN International Conference on the Peaceful Uses of Atomic Energy, Geneva, A/CONF 28/P/538 (1964).
Reproduced with permission from Kirk Sorensen (http://thoriumenergy.blogspot.com/2006/06/1972-summary-of-ornl-fluoride-reactor.html)