Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets
Journal Article Patterson-Hine, F; Davidson, J; Klein, D; ... - J. Fusion Energy; [United States]
Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs: fluorination only, fluorination plus reductive extraction, and fluorination, plus reductive extraction, plus metal transfer. The effects of processing on blanket performance have been assessed for these three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis, which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The method of salt processing was found to have little affect on the level of radioactivity, toxicity, or the thermal behavior of the salt during operation of the reactor. The processing rates necessary to maintain the desired uranium concentrations in the suppressed-fission environment were quite low, which permitted only long-lived species to be removed from the salt. The effects of the processing therefore became apparent only after the radioactivity due to the short-lived species diminished. The effect of the additional processing [reductive extraction and metal transfer] could be seen after approximately 1 year of decay, but were not significant at times closer to shutdown. The reduced radioactivity and corresponding heat deposition were thus of no consequence in accident or maintenance situations. Net fissile production in the Be/MS blanket concept at a fusion power level of 3000 MW at 70% capacity ranged from 5100 kg/year to 5170 kg/year for uranium concentrations of 0.11% and 1.0% /sup 233/U in thorium, respectively, with fluorination-only processing. The addition of processing by reductive extraction resulted in 5125 kg/year for the 0.11% /sup 233/U case and 5225 kg/year for the 1.0% /sup 233/U case.
- //doi.org/10.1007/BF01082702
NUCLEAR CHARACTERISTICS OF SPHERICAL, HOMOGENEOUS, TWO-REGION, MOLTEN- FLUORIDE-SALT REACTORS
Technical Report Alexander, L.; Carrison, D.; MacPherson, H.; ...
ABS>The use of a molten-salt fuel makes possible the production of hihh- pressure, superheated steam with a nuclear reactor operating at low pressure. The corrosion resistance of the INOR-8 series of nickelmolpbdenum alloys appears to be sufficient to guarante reactor component life-times of 10 to 20 years. Proposed continuous fuel-processing methods show promise of reducing fuel- processing costs to negligible levels. With U/sup 233/ as the fuel and Th/sup 232/ as the fertile material in both core and blanket, initial regeneration ratios up to 1.08 can be obtained at critical masses less than 600 kg. The corresponding inventory for a 600-Mw[th] central station power reactor is initially about 1300 kg. With U/sup 235/ as the fuel, U/sup 233/ is produced, and initial regeneration ratios in excess of 0.6 can be obtained with cnitical masses of less than 300 kg. The corresponding critical inventories for 600- Mw[th] central station power reactors are 600 kg or less, depending on the thorium loading. It is concluded that homogeneous, molten-salt-fueled reactors are competitive in regard to nuclear performance with present solid-fuel reactors, and they may be oconomically superior because of lower fuel and fuel- processing costs. [auth]
- //doi.org/10.2172/4200482
- Full Text Available
Gaseous fission product management for molten salt reactors and vented fuel systems
Conference Messenger, S.; Forsberg, C.; Massie, M.
Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors [MSRs] and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton [e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years]. Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF [78 mole percent] - [HN]F 4 [22 mole percent] with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options for disposal of fission gas wastes. In each option, lithostatic pressure, a kilometer or more underground, eliminates the pressure driving force for noble gas release and dissolves any untrapped gas in deep groundwater or into incorporated solid waste forms. The options, challenges, and potential for these methods to dispose of gaseous fission products are described. With this research, we hope to help both MSRs and other advanced reactors come one step closer to commercialization. [authors]
Magnetic mirror fusion-fission early history and applicability to other systems
Technical Report Moir, R
In the mid 1970s to mid 1980s the mirror program was stuck with a concept, the Standard Mirror that was Q {approx} 1 where Q=P{sub fusion}/P{sub injection}. Heroic efforts were put into hybridizing thinking added energy and fuel sales would make a commercial product. At the same time the tokamak was thought to allow ignition and ultrahigh Q values of 20 or even higher. There was an effort to use neutral beams to drive the tokamak just like the mirror machines were driven in which case the Q value plunged to a few, however this was thought to be achievable decades earlier than the high Q versions. Meanwhile current drive and other features of the tokamak have seen the projected Q values come down to the range of 10. Meanwhile the mirror program got Q enhancement into high gear and various tandem mirrors projected Q values up towards 10 and with advanced features over 10 with axi-symmetric magnets [See R. F. Post papers], however the experimental program is all but non-existent. Meanwhile, the gas dynamic trap mirror system which is present day state-of-the-art can with low risk produce Q of {approx}0.1 useful for a low risk, low cost neutron source for materials development useful for the development of materials for all fusion concepts [see Simonen white paper: 'A Physics-Based Strategy to Develop a Mirror Fusion-Fission Hybrid' and D.D. Ryutov, 'Axisymmetric MHD-stable mirror as a neutron source and a driver for a fusion-fission hybrid']. Many early hybrid designs with multi-disciplinary teams were carried out in great detail for the mirror system with its axi-symmetric blanket modules. It is recognized that most of these designs are adaptable to tokamak or inertial fusion geometry. When Q is low [1 to 2] economics gives a large economic penalty for high recirculating power. These early studies covered the three design types: Power production, fuel production and waste burning. All three had their place but power production fell away because every study showed fusion machines that were extensively studied by multidisciplinary teams came up with power costs much higher than for existing fission plants except in very large sizes [3 GWe]. There was lots of work on waste burning - Ted Parrish - comes to mind. However, fuel production along with power production became nearly everyone's goals. First, fast-fission blankets were favored but later to enhance safety, fission-suppressed blankets came into vogue. Both fuel producing and waste burning hybrid studies were terminated with the advent of accidents, high interest rates, rising 'green like' movement and cheap natural gas for power production. For waste burning and fast-fission fuel producing designs, the blanket energy multiplication was about 10 and economics was OK relative to recirculating power for Q over 2. For fission-suppressed fuel producers, where the blanket multiplication is under 2, the Q needed was over 5. In the mirror program we came at this problem by trying to find a product for mirror fusion technology. We hoped we had a product and studied and promoted it. There was no market pull and when the mirror program collapsed in the US, so did both hybrid programs for mirrors and tokamaks and IFE by the mid 1980s. Today, the problem of what to do with wastes that were supposed to be accepted by the government appears to be a high value market pull. It remains to be shown if fusion neutrons can be generated at low enough cost so that economics will not be a showstopper. For burning only the minor actinides, the economics will be the most favorable. Burning the Pu as well will lower the number of fission reactors supported by each burner fusion machine and hurt economics of the system. The fuel-producing role of fusion to fuel fission reactors remains an important possible use of fusion especially in the early stages of fusion development. It is not clear that burning fission wastes in a fusion machine is more appropriate than burning these wastes in specially designed fission machines. Fusion can produce U-233 along with over 2.4%U-232 making the material largely nonproliferating and this material can in effect add neutrons to a fission reactor that would otherwise be short of reactivity to burn wastes. Similar ideas apply to Pu production. Unlike enrichment, producing U-233 does not burden the system with lots of U-238 with its source of more actinide wastes. The idea is fission plants are already designed and proven to fission at impressive power density and safety whereas fusion machines will have a harder time showing workability with thin walls separating the awkward geometry of the high curie inventory from the vacuum chamber that will get lots of radiation damage.
- //doi.org/10.2172/964097
- Full Text Available
MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JANUARY 31, 1958
Technical Report
Reactor Design Studies. Preliminary layouts were prepared of the reactor system based on the use of five fuel pumps and two blanket-salt pumps. Total power is assumed to be 638 Mw. Parameter studies of tworegion molten salt reactors were continued through nuclear calculations on the UNIVAC, Oracle, and IBM704. Comparisons of Li--Be and Na--Be salts showed that the Na--Be salt requires 1.2 to 2 times the U/sup 235/ inventory required by Li --Be. A preliminary fill-anddrain system design was completed which satisfies design criteria. The system vessel consists of 48 20-ft lengths of 12-in.-diam pipe arranged in 6 vertical banks. A hydrostatic bearing was designed for use in fuel pumps which differs from conventional bearings in that the pockets rotate on the impeller. Pump design studies established the characteristics of the five fuel pumps for use in the BeF/sub 2/--LiF--UF/sub 4/ fueled reactor. Studies of pump arrangement and salt-lubricated pump bearings were continued. Further developments in remote maintenance techniques and testing of spectal flanges are reported. Heat exchanger tests are being conducted to determine the heat transfer characteristics of various salts in contact with a number of structural materials. Additional dynamic corrosion studies of INOR-8 and Inconel in fuel salts were conducted. Three proposed entrance-exit systems are being evaluated through the use of glass models; one straight-through system and two concentric- flow systems. Studies of the properties of salt systems having various compositions were continued. Preliminary studies were made of naturalconvection molten-salt reactors. Comparison with forced-circulation reactors of 60-Mw indicated a 42% greater fuel inventory is required in the convection reactor. Materials Studies. Studies are under way for determining the effect of carburization on the properties of reactor structural materials. Precious-metal brazing alloys being considered for use in fused salt heat exchangers were corrosion tested in fuel mixtures. Elasticity, thermal conductivity, and tensile property measurements were made on air-melted heats of INOR8. The sorptive properties of charcoal for Xe were significantly improved by heat treatment. Phase equilibrium studies were made of ThF/sub 4/ and UF/sub 4/ in mixtures of LiF--BeF/sub 2/. The solubility of PuF/sub 3/ in Becontaining fluoride salts was investigated. A survey was made of the properties of fused chlorides of interest as secondary heat transfer fluids. LiCl --RbCl appears most attractive. Solubility studies were made of noble gases in NaF-KF --LiF, HF in NaF--ZrF/sub 4/ , and fission product fluorides in NaF--BeF/sub 2/ and LiF -BeF/sub 2/. Precipitation of fission product oxides from fluoride melts was also investigated. [For preceding period see ORNL-2431.] [D.E.B.]