The basic purpose of ORNL-4528 differed from other MSR designs between 1962 and 72. Unlike other reactor system design projects ORNL-4528 was not written to as a part of an ongoing development program. Rather it was written after the two fluid line of development it represented had been dropped in favor of a single fluid design. ORNL-4528 was one of five 1 GWe MSR designs developed between 1961 and 1971 by ORNL or by associated engineering firms. The purpose of the other 4 designs was explained by ORNL-5018:
The objectives of this activity are: (I) to develop the conceptualBecause the line of research document led by ORNL-4528, its intent was not to offer clues for future development, but to document a terminated line of research. Many ORNL scientists, including my father, were not in agreement with the decision to abandon the two fluid approach, their continued believe in the soundness of their views, may have motivated the desire to document the modular two fluid design.
design for a commercial 1000 MW(e) MSBR in sufficient detail to identify the major areas im which additional technology development is required and to produce meaningful estimates of the nuclear and economic performances of this reactor type, (2) to develop the design criteria and conceptual design for a molten-salt demonstration reactor that will provide the information necessary for construction of commercial MSBRs in sufficient detail to identify additional technology development which is required for construction of the demonstration reactor and to provide improved estimates of the capital and operating costs for the demonstration reactor, (3) to develop the design criteria and conceptual design for a molten-salt test reactor in sufficient detail to identify additional
technology development which is required for construction of the test reactor and to provide improved estimates of the capital and operating costa for the test reactor, and (4) to develop the design criteria and conceptual design for a molten-salt teat reactor mockup in sufficient detail to identify additional technology development which is required for construction of the test reactor mockup and to provide improved estimates of the capital and operating costs for the mckup.
An additional important objective of this activity is the examination of alternate reactor types such as molten-salt converter reactors using uranium or plutonium fuel makeup as well as uses for molten-salt reactors other than large central station electric power generation in sufficient detail to assess the likely economic importance of alternate molten-salt reactor types. Limited conceptual design work would be carried out on alternate reactor types which show promise.
At any rate the design documented by ORNL-4528 is far from mature and contains flaws. I would encourage readers to find flaws and comment on them.
UC-80 - Reactor Technology
TWO-FLUID MOLTENSALT BREEDER REAmOR DESIGN STUDY
(STATUS AS OF JANUARY 1, 1968)
R. C. Robertson
R. B. Briggs
O. L. Smith
E. S. Bettis
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-U-233 fuel cycle could produce electric power at sufficiently low cost to be of interest and at the same time show good utilization of U.S. nuclear fuel resources. This report covers the progress made in the study up to August 1967, at which time the two-fluid MSBR work was set aside in order to study a single-fluid MSBR concept. The latter became of interest at that time due to the discovery that protactinium and other fission products could be separated from a uranium-and-thorium-bearing fuel salt by reductive extraction into liquid bismuth.
The two-fluid MSBR is graphitemoderated and -reflected, with a 'LiF-BeFz-UFe fuel salt circulated through the core and a 'LiF-ThF4-BeF2 blanket salt circulated through separate flow channels distributed throughout the core, as well as in a surrounding under moderated region. The fissings raise the 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 shell-and-tube heat exchangers to raise the temperature of a circulating NaBF4-NaF coolant salt to about 1150°FbThe co$ant salt transports the heat to steam generators and reheaters to provide 3500-psia 1000 F/l000 F steam for a conventional turbine generator.
The conceptual design was based on use of four reactors and the associated heat transfer systems in a socalled modular arrangement to supply steam to a single turbine-generator. This made it practical to consider replacement of an entire reactor vessel assembly after the core graphite received its allowable exposure to neutrons. The total fluence at which it was thought that additional graphite dimensional changes would become excessive was taken as 3 x neutrons/cm2 (E > 50 kev), or about eight years of full-power operation.
All portions of the systems in contact with the fluoride or fluoroborate salts would be fabricated of Hastelloy N that has a small amount of titanium added to improve 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 material. The two-fluid concept involves joining graphite core elements to Hastelloy N tubing using a brazing process developed at ORNL.
The reactors and associated systems would be housed in concrete cells to provide biological shielding and double containment of all 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 no formal safety analysis was completed.
Fuel and blanket salts would be continuously processed in a nearby cell 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 rareearth fission products from the core on a 6-y cycle. Since no conceptual designs for the chemical plant were completed, cost estimates could not be on a definitive basis. The tentatively estimated fuel cycle cost is about 0.5 mill/kwhr, which includes the fixed charges and operating costs for the processing equipment, the fuel inventory charge, and the credit for bred fuel. Graphite replacement costs, which are not included, would add about 0.2 mill/kWhr.
The tentatively estimated total construction cost of a 1ooo-Mw(e) MSBR station, based on the early 1968 value of the 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, fuel processing, afterheat removal, and maintenance systems needed further investigation at the time the study was suspended, and the limited performance 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 two- fluid molten-salt reactor power statiohThe major concern was whether mechanical failure of graphite tubes in the reactor core would cause the effective lifetime of the core to be significantly less than the eight years imposed by the effects of irradiation on the graphite.
Update 4/16/09: Kirk is publishing sections of ORNL-4528 on Energy from Thorium.