ORNL's Molten-Salt Reactor Program (1958-1976)

The objective of the Molten-Salt Reactor Program (MSRP) was the development of nuclear reactors that used fluid fuels that are solutions of fissile and fertile materials in suitable carrier salts. The program was an outgrowth of the effort begun in the early 1950s in the Aircraft Nuclear Propulsion (ANP) Program to make a fluid-fueled reactor power plant for aircraft. A liquid-fluoride reactor, the Aircraft Reactor Experiment, was operated at Oak Ridge National Laboratory in 1954 as part of the ANP Program.

The major goal of the MSRP was to achieve a thorium reactor that could produce power at low cost while simultaneously conserving and extending the nation’s fuel resources. Fuel for this type of reactor would be 233UF4 dissolved in a mixture of LiF and BeF2, but 235U or plutonium could be used for startup. The fertile material would be ThF4 dissolved in the same salt of in a separate blanket salt of similar composition. The technology being developed for the thorium reactor was also applicable to high-performance converter reactors.

A major program activity from 1965 through 1969 was the operation of the Molten-Salt Reactor Experiment (MSRE). This reactor was built to test the types of fuels and materials that would be used in thorium reactors; it also provided operation and maintenance experience. The MSRE operated at 650°C and produced 7.3 MW of heat. The initial fuel contained 0.9 mole% UF4, 5% ZrF4, 29% BeF2, and 65% 7LiF; the uranium was about 33% 235U. The fuel circulated through a reactor vessel and an external pump and heat exchange system. Heat produced in the reactor was transferred to a coolant salt, and the coolant salt was pumped through a radiator to dissipate the heat to the atmosphere. All this equipment was constructed of Hastelloy N, a nickel-molybednum-iron-chromium alloy. The reactor contained an assembly of graphite moderator bars in direct contact with the fuel.

Design of the MSRE was started in 1960, fabrication of equipment began in 1962, and the reactor became critical on June 1,1965. Operation at low power began in January 1966, and sustained power operation was begun in December 1966. One run continued for six months, until stopped on
schedule in March 1968.

Completion of this six-month run ended the first phase of MSRE operation, in which the objective was to show, on a small scale, the attractive features and technical feasibility of these systems for commercial
power reactors. The conclusion was that this objective had been achieved and that the MSRE had shown that liquid-fluoride reactors can be operated at 650°C without corrosive attack on either the metal or graphite parts of the system; also the fuel is stable; the reactor equipment can operate satisfactorily at these conditions; xenon can be removed rapidly from liquid salts; and when necessary, the radioactive equipment can be repaired or replaced.

The second phase of MSRE operation began in August 1968, when a small facility in the MSRE building was used to remove the original uranium charge from the fuel salt by treatment with gaseous F2. In six days of fluorination, 221 kg of uranium was removed from the molten salt and loaded into absorbers filed with sodium fluoride pellets. The decontamination and recovery of the uranium were very good.

After the fuel was processed, a charge of 233U was added to the original carrier salt, and in October 1968 the MSRE became the world’s first reactor to operate on 233U. The nuclear characteristics of the MSRE with the 233U were close to the predictions, and the reactor was quite stable. In September 1969, small amounts of PuF3 were added to the fuel to obtain some experience with plutonium in a molten-salt reactor. The MSRE was shut down permanently December 12, 1969, so that the funds supporting its operation could be used elsewhere in the research and development program.

Because of limitations on the chemical-processing methods available in the past, most of the work on thorium reactors was aimed at two-fluid systems in which graphite tubes would be used to separate uranium-bearing fuel salts from thorium-bearing fertile salts. However, in late 1967 a one-fluid thorium was thought to be feasible with the development of processes that use liquid bismuth to isolate protactinium and remove rare earth fission products from a salt that also contains thorium. ORNL MSRP studies showed that a one-fluid thorium reactor based on these processes can have fuel-utilization characteristics approaching those of their two-fluid design concepts. Since the graphite serves only as moderator, the one-fluid reactor was more nearly a scale-up of the MSRE. These advantages caused a change in the emphasis of the program from the two-fluid to the one-fluid thorium reactor; most of the design and development effort through the end of the program was directed to the one-fluid system.

In the congressional authorization report on the United States Atomic Energy Commission’s (USAEC) programs for FY1973, the Joint Committee on Atomic Energy recommended that the molten-salt reactor be appraised so that a decision could be made about its continuation and the level of funding appropriate for it. Consequently, a thorough review of fluoride reactor technology was undertaken to provide information for an appraisal. A significant result of the review was the preparation of ORNL-4812, The Development Status of Molten-Salt Breeder Reactors (34.2MB PDF). This led to the issue of a report from the USAEC, WASH-1222, An Evaluation of the Molten-Salt Breeder Reactor (4.4MB PDF). A subsequent decision was made by the USAEC to terminate work on liquid-fluoride reactors for budgetary reasons; in January 1973 ORNL was directed to conclude fluoride reactor work.

In January 1974, the USAEC program for liquid-fluoride reactor development was reinstated. A considerable effort during 1974 was concerned with assembling a program staff, making operational a number of development facilities used previously, and replacing a number of key developmental facilities that had been reassigned to other reactor programs. A significant undertaking was the formulation of detailed plans for the development of liquid-fluoride thorium reactors and the preparation of ORNL-5018, Program Plan for Development of Molten-Salt Breeder Reactors (49.6MB PDF).

During 1974 and 1975, work in the Molten-Salt Reactor Program was devoted to the technology needed for molten-salt reactors. The work included conceptual design studies and work on materials, the chemistry of fuel and coolant salts, fission-product behavior, processing methods, and the development of systems and components. The most important single aspect of the program was work on the development and demonstration of an alloy that is suitable for the primary circuit of an MSBR and has adequate resistance to tellurium-induced shallow intergranular cracking, which was first observed in MSRE surveillance specimens. A second important area consisted of studies of the chemical interaction of tritium with the MSBR secondary coolant, both in laboratory chemistry studies and in a large engineering facility (Coolant Salt Technology Facility). These studies culminated in the demonstration of an adequate basis for management of tritium in a 1000 MWe thorium reactor.

In February 1976, ORNL was directed by Energy Research and Development Administration (ERDA), the successor to the AEC, to again terminate the Molten-Salt Reactor Program for budgetary reasons. Work during the remainder of FY 1976 was directed toward completion of short-term work in the Program, reporting of associated information, and the assignment of the MSRP staff and experimental facilities to other ORNL programs.

The results of the Molten-Salt Reactor Program were documented in hundreds of papers, and specifically in a series of semiannual progress reports, stretching from 1958 to 1967. The tables-of-contents from all of these reports are available in a single, compact PDF document.

Molten-Salt Reactor Program Progress Reports Tables-of-Contents (PDF, 362KB)

Furthermore, specific sections from some of the key reports have been extracted and presented in page format on this site. Follow these links for more information.

ORNL-4119, MSRP Semiannual Progress Report for Period Ending February 28, 1967

ORNL-4191, MSRP Semiannual Progress Report for Period Ending August 31, 1967

ORNL-4254, MSRP Semiannual Progress Report for Period Ending February 29, 1968

ORNL-4344, MSRP Semiannual Progress Report for Period Ending August 31, 1968

ORNL-4396, MSRP Semiannual Progress Report for Period Ending February 28, 1969