ISSN:0550-3043    

DOI:10.13182/NT70-A28617

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

Molten-salt breeder reactors (MSBR’s) are being developed at the Oak Ridge National Laboratory because of their promise for generating low-cost power while conserving and extending our resources of fissionable uranium. The circulating liquid fuel makes MSBR’s much different from present power reactors, but this type fuel provides the potential fqr reducing both power costs and the amount of uranium that must be mined to fuel the nuclear power industry.

ISSN:0550-3043    

DOI:10.13182/NT70-A28618

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

Molten-salt breeder reactors (MSBR’s) are being developed by the Oak Ridge National Laboratory for generating low-cost power while extending the nation’s resources of fissionable fuel. The fluid fuel in these reactors, consisting of UF4and ThF4dissolved in fluorides of beryllium and lithium, is circulated through a reactor core moderated by graphite. Technology developments over the past 20 years have culminated in the successful operation of the 8-MW(th) MoltenSalt Reactor Experiment (MSRE), and have indicated that operation with a molten fuel is practical, that the salt is stable under reactor conditions, and that corrosion is very low. Processing of the MSRE fuel has demonstrated the MSR processing associated with high-performance converters. New fuel processing methods under development should permit MSR’s to operate as economical breeders. These features, combined with high thermal efficiency (44%) and low primary system pressure, give MSR converters and breeders potentially favorable economic, fuel utilization, and safety characteristics. Further, these reactors can be initially fueled with233U,235U, or plutonium. The construction cost of an MSBR power plant is estimated to be about the same as that of light-water reactors. This could lend to power costs ~0.5 to 1.0 mill/kWh less than those for light-water reactors. Achievement of economic molten-salt breeder reactors requires the construction and operation of several reactors of increasing size and their associated processing plants.

ISSN:0550-3043    

DOI:10.13182/NT70-A28619

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

The MSRE is an 8-MW(th) reactor in which molten fluoride salt at 1200°F circulates through a core of graphite bars. Its purpose was to demonstrate the practicality of the key features of molten-salt power reactors.Operation with235U (33% enrichment) in the fuel salt began in June 1965, and by March 1968 nuclear operation amounted to 9000 equivalent full-power hours. The goal of demonstrating reliability had been attained—over the last 15 months of235U operation the reactor had been critical 80% of the time. At the end of a 6-month run which climaxed this demonstration, the reactor was shut down and the 0.9 mole% uranium in the fuel was stripped very efficiently in an on-site fluorination facility. Uranium-233 was then added to the carrier salt, making the MSRE the world's first reactor to be fueled with this fissile material. Nuclear operation was resumed in October 1968, and over 2500 equivalent full-power hours have now been produced with233U.The MSRE has shown that salt handling in an operating reactor is quite practical, the salt chemistry is well behaved, there is practically no corrosion, the nuclear characteristics are very close to predictions, and the system is dynamically stable. Containment of fission products has been excellent and maintenance of radioactive components has been accomplished without unreasonable delay and with very little radiation exposure.The successful operation of the MSRE is an achievement that should strengthen confidence in the practicality of the molten-salt reactor concept.

ISSN:0550-3043    

DOI:10.13182/NT8-2-118

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

This document summarizes the large program of chemical research and development which led to selection of fuel and coolant compositions for the Molten-Salt Reactor Experiment (MSRE) and for subsequent reactors of this type. Chemical behavior of the LiF-BeF2-ZrF4-UF4fuel mixture and behavior of fission products during power operation of MSRE are presented. A discussion of the chemical reactions which show promise for recovery of bred233Pa and for removal of fission product poisons from a molten-salt breeder reactor is included.

ISSN:0550-3043    

DOI:10.13182/NT70-A28621

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

Operating experience with the Molten-Salt Reactor Experiment (MSRE) has demonstrated the excellent compatibility of the graphite-Hastelloy-N-fluoride salt system at 650°C. Several improvements in materials are needed for a molten-salt breeder reactor with a basic plant life of 30 years; specifically: Hastelloy-N with improved resistance to embrittlement by thermal neutrons; graphite with better dimensional stability in a fast neutron flux; graphite that is sealed to obtain a surface permeability of<10−8cm2/sec; and a secondary coolant that is inexpensive and has a melting point of ~400°C. A brief description is given of the materials work in progress to satisfy each of these requirements.

ISSN:0550-3043    

DOI:10.13182/NT70-A28622

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

The molten-salt breeder reactor being developed at Oak Ridge National Laboratory (ORNL) requires continuous chemical processing of the fuel salt,7LiF-BeF2-ThF4(72-16-12 mole%) containing ~0.3 mole%233UF4. The reactor and the processing plant are planned as an integral system. The main functions of the processing plant will be to isolate233Pa from the neutron flux and to remove the rare-earth fission products. The processing method being developed involves the selective chemical reduction of the various components into liquid bismuth solutions at ~600°C, utilizing multistage counter-current extraction. Protactinium, which is easily separated from uranium, thorium, and the rare earths, would be trapped in the salt phase in a storage tank located between two extraction contactors and allowed to decay to233U. Rare earths would be separated from thorium by a similar reductive extraction method; however, this operation will not be as simple as the protactinium isolation step because the rare-earth-thorium separation factors are only 1.3 to 3.5. The proposed process would employ electrolytic cells to simultaneously introduce reductant into the bismuth phase at the cathode and to return extracted materials to the salt phase at the anode. The practicability of the reductive extraction process depends on the successful development of salt-metal contactors, electrolytic cells, and suitable materials of construction.

ISSN:0550-3043    

DOI:10.13182/NT70-A28623

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

Existing data an dimensional changes in graphite have been fitted to parabolic temperature-sensitive curves. From these, the graphite life, radiation-induced stresses, and permissible geometries have been calculated. It is concluded existing materials can be utilized in a molten-salt reactor which has a core graphite life of about four years, without serious cost penalty.Fission product xenon can be removed by sparging the fuel salt with helium bubbles and removing them after enrichment. With reasonable values of salt-to-bubble transfer coefficient and graphite permeability, the penalty to breeding ratio can be reduced to<0.5%.

ISSN:0550-3043    

DOI:10.13182/NT70-A28624

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

A conceptual design has been made of a single-fluid 1000 MW(e) Molten-Salt Breeder Reactor (MSBR) power station based on the capabilities of present technology. The reactor vessel is ~22ft in diameter×20 ft high and is fabricated of Hastelloy-N with graphite as the moderator and reflector. The fuel is233U carried in a LiF-BeF2-ThF4mixture which is molten above 930°F. Thorium is converted to233U in excess of fissile burnup so that bred material is a plant product. The estimated fuel yield is 3.3% per year.The estimated construction cost of the station is comparable to PWR total construction costs. The power production cost, including fuel-cycle and graphite replacement costs, with private utility financing, is estimated to be 0.5 to 1 mill/kWh less than that for present-day light-water reactors, largely due to the low fuel-cycle cost and high plant thermal efficiency.After engineering development of the fuel purification processes and large-scale components, a practical plant similar to the one described here appears to be feasible.

ISSN:0550-3043    

DOI:10.13182/NT70-A28625

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

As presently conceived at Oak Ridge National Laboratory and described in this issue, the single-fluid Molten-Salt Breeder Reactor, operating on the232Th-233U fuel cycle and based on a reference design, has a breeding ratio of ~1.06, specific fissile inventory of 1.5 kg/MW(e), a fuel doubling time of ~20 years, and fuel cycle costs of ~0.7 mill/kWh(e). Start-up may be accomplished with either enriched uranium or plutonium, with little effect on fuel cost; the breeding ratio, averaged over reactor life, is reduced 0.01 to 0.02 relative to the equilibrium cycle.Operated as a converter, with limited chemical processing, the reactor may have a conversion ratio in the range 0.8 to 0.9 with fuel cycle costs of 0.7 to 0.9 mill/kWh(e).

ISSN:0550-3043    

DOI:10.13182/NT70-A28626

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor