TWO-FLUID MOLTEN-SALT BREEDER REACTOR DESIGN STUDY (STATUS AS OF JANUARY 1, 1968). (1970)

A conceptual design study of a 1000-Mw(e) thermal breeder power station based on a two-fluid MSBR was commenced in 1966 as part of a program to determine whether a molten-salt reactor using the thorium-233 U fud cycle could produce ekctr.cpower at sufficiently k>w cost to be of interest and at the same time show good utilization of U.S. nucfcsar fuel resources.This report covers the progress made in the study up to August 1967, 2t vhich fi 1 ™* th* twv»-flw<?MSBR work w» set aside in order to study a single-fluid MSBR concert.The atter became of interest at that time due to the discovery that protactinium and other fission products could be separated from a uraiimro-and-thorium-beanng fuel salt by reductive extraction into liquid bismuth.The two-fluid MSBR is graphite-moderated and -reflected, with a 7 LiF-BeF 2 -UF 4 fuel salt circulated through the core and a 7 LiF-ThF 4 -BeFj blanket salt ckculated through separate flow channels distributed throughout the core, as weO as in a surrounding undermoderjted region.The fissions raise die temperature of the fuel salt to about 1300°F and that of the blanket salt to about 1250°F.Heat is removed from the salts in shelUnd-tube heat exchangers to raise the temperature of & circulating NaBF 4 -NaF coolant salt to about 1150 F. The coolant salt transports the heat to steam generators and reheaters to provide 3500-psia 1000°F/1U00°F steam for a ooventional turbaiegenerator.The conceptual design was based en use of four wuctorr rsd the «MOci*icu heai oansfer sy.tems in a so-cailed modular arrangement to supply steam to a single turbine-generator.This made it practical to consider replacement of as entire react*?vessel assembly after the core graphite received its allowable exposure to neutrons.The total fhience at which it was thought that additional graphite dimensional changes would become excessive w«s takes «3X iO 22 neutrons/cm 3 (E > 50 kev), or about eight years of fun-power operation.All portions of the systems in contact with the fluoride or flocroborate salts would be fabricated of HasteUoy N that har 2 small amount of titanium added to iraptov-the resistance to radiation damage.The graphite would be a specially coated grade having low gas permeability to xenon and better resistance to radiation damage than conventional mataiaL The two-fluid concept involves joining graphite core dements to HasteUoy N tubing using a brazing process developed at ORNL.The reactor* and associated systems would be housed in concrete cefls to provide biological shielding and double containment of aO radioactive materials.Plant flowsheets and layouts were developed sufficiently during the study to give an indication of feasibility and to give a basis for cost estimates, but no optimization studies were made.Safety aspects were considered throughout the design effort, but ao fonnai safety analysis was completed.Fud and blanket salts would be continuously processed in a nearby cdl to remove fission products and to recover the bred product.The processing rate would correspond to removal of uranium and protactinium from the blanket on a 3-day cycle and rare-earth fission products from the core on a 60-day cycle.Since no conceptual designs for the chemical plant were completed, cost estimates could not be on a definitive basis.The tentatively estimated fud cycle cost is about 0 J nuH/kwhr, which includes the fixed charges and operating costs for the processing equipment, the rod inventory charge, and the credit for bred fuel.Graphite replacement costs, which are not included, would add about 0.2 mill/kwhr.The tentativdy estimated total construction cost of a 1000-Mw(e) MSBR station, based on the early 1968 value of die dollar, is about $141 per kilowatt.The power production cost for a privately owned station, based on fixed charges of 13.7% and 80% plant factor, is about 4 mills/kwhr.The net thermal efficiency of the plant would be about 44.9%.The off-gas, fud processing, afterheat removal, and maintenance systems needed former investigation at the time die study was suspended, and the limited perfor-.uce of the graphite undoubtedly restricts the design and imposes a maintenance penalty, but the study did not disclose any aspects which indicated that major technological discoveries would be required to design a twofluid molten-salt reactor power station.The major concern was whether mechanical failure of graphite tubes in the reactor core would cause the effective lifetime of the cere to be significantly less than the eight years imposed by the effects of irradiation on the graphite.1 2 INTRODUCTIONThe basic objective of the Molten-Salt Reactor Pro gram is to develop the technology for economical nuclear power reactors that make use of fluid fuels which are solutions of fissile and fertile materia!? in suitable carrier salts.A major goal is to achieve a thermal breeder reactor based on the tliorium-2 ' 3 U fuel cycle that will produce power at low cost while conserving and extending the nation's fuel resources.Conceptual design studies of a variety of molten-salt breeder reactors for large plants are sn important part of this program.In August 1966 we published a survey report, ORNL-3996, 1 in which we described briefly the status of molten-salt reactor technology and the designs of reactors and fuel processing facilities for 1000-Mw(e) power stations.This survey led us to conclude that the two-fluid reactor which separates the fuel and blanket salts held the most promise for development as a breeder reactor.The modular version, consisting of four reactor modules and associated intermediate systems supplying steam to one turbine-generator, was selected for more detailed analysis.The study of the modular design of a 1000-Mw(e) plant was begun in the fall of 1966, and some of the results were published in the MSRP progress reports, ORNL4037, 2 ORNL4119, 3 and 0RNL419L 4 Much of the effort was spent on designs for the core and in exploring the effects of radiation-induced damage to graphite on the core designs.The plant layout, the cell designs, the drain tank systems, the nuclear character istics, the maintenance, and the cost estimates were also examined in more detail than had been possible in the earlier survey.Considerable progress had been made in these studies when, in August 1967, encouraging information ob tained frcrn research on the processing oi mulicn-sali fuels indicated that protactinium and some fission products could be separated from the uranium-andthoriunvcontaining fuel salt of a ont-fluid reactor by reductive extraction into liquid bismuth.At about this same time, nuclear calculations indicated that a conver sion ratio greater than 1 could be achieved in a 'Fuil R.