Molten-Salt Reactor Experiment
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The Molten-Salt Reactor Experiment (MSRE) was an experimental molten-salt reactor at the Oak Ridge National Laboratory (ORNL), which took the lead in researching this technology through the 1960s.
The MSRE was a 7.4 MWth test reactor located in the Melton Valley area at ORNL. Its piping and structural components were made from Hastelloy-N, and its core was graphite. It went critical in 1965 and ran for four years. The fuel for the MSRE was LiF-BeF2-ZrF4-UF4 (65-30-5-0.1), the graphite core moderated it, and its secondary coolant was FLiBe (2LiF-BeF2). It operated at a peak temperature of 650 °C and operated for the equivalent of about 1.5 years of full power operation.
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[edit] Purpose
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. The purpose of the Molten-Salt Reactor Experiment was 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.
[edit] Description
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. Haubenreich and Engel[1], Robertson[2], and Lindauer[3] provide more detailed descriptions of the reactor and processing plant.
[edit] 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 "MSR Program Semiannual Progress Report" [4].)
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 [[4], 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 2¼-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 basis 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 U.S. Atomic Energy Commission-owned shops at Oak Ridge and Paducah [[4], 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 [5].)
An extensive out-of-pile corrosion test program was carried out for Hastelloy-N [[4], 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 100 °C [[4], pp. 252–257].
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 [[4], 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 [6]. The primary systems were installed by ORNL forces; subcontractors modified the building and installed ancillary systems.
[edit] Operation
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 [1] and [7], 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 [8], 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[3]. 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.
[edit] Results
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 [7]. 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 [9]. 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.
[edit] Decommissioning
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[edit] External links
[edit] References
- ^ a b P.N. Haubenreich and J.R. Engel (1970). "Experience with the Molten-Salt Reactor Experiment" (PDF, reprint). Nuclear Applications and Technology 8: 118–136.
- ^ R.C. Robertson (January 1965). "MSRE Design and Operations Report, Part I, Description of Reactor Design". ORNL-TM-0728.
- ^ a b R.B. Lindauer (August 1969). "Processing of the MSRE Flush and Fuel Salts". ORNL-TM-2578.
- ^ a b c d e f (July 31, 1964) "MSR Program Semiannual Progress Report". ORNL-3708.
- ^ H.E. McCoy et al. (1970). "New Developments in Materials for Molten-Salt Reactors". Nuclear Applications and Technology 8: 156.
- ^ B.H. Webster (April 1970). "Quality-Assurance Practices in Construction and Maintenance of the MSRE". ORNL-TM-2999.
- ^ a b M.W. Rosenthal, P.N. Haubenreich, H.E. McCoy, and L.E. McNeese (1971). "Current Progress in Molten-Salt Reactor Development". Atomic Energy Review IX: 601–50..
- ^ B.E. Prince, S.J. Ball, J.R. Engel, P.N. Haubenreich, and T.W. Kerlin (February 1968). "Zero-Power Physics Experiments on the MSRE". ORNL-4233.
- ^ R.B. Briggs (Winter 1971–72). "Tritium in Molten-Salt Reactors". Reactor Technology 14: 335–42.
