Molten Salt Breeder Reactor Research Articles

The oxide chemistry of protactinium in molten fluorides

The solubility behavior of an insoluble oxide phase of protactinium(V) has been studied in molten LiFBeF 2ThF 4 (72-16-12 mole%)-the proposed solvent salt for the Molten Salt Breeder Reactor (MSBR)-as a function of temperature and redox potential. The stoichiometry of the precipitation indicated the 0 Pa ratio to be 2·5 in this oxide phase; however, it could not be identified as Pa 2O 5. Moreover, the apparent entropy of solution of the precipitate, as indicated by the high temperature dependence of its solubility, is higher than expected for Pa 2O 5 by ∼23 entropy units per mole of protactinium. For these reasons, this oxide phase is thought to be an addition compound PaO 2·5. nLiF. It has such a low solubility − log(X PaF 5 ·X O 2− 5 2 ) = 0·91-12·76 ( 10 3 T ) —that the PaF 5 concentration in solution can be reduced by the addition of oxide to ∼ 1 ppm at 550°C before ThO 2 saturation is reached. With UF 4 present at concentrations typical of an MSBR fuel (0·3 mole%), the PaF 5 concentration can be lowered to ∼ 10 ppm (550°C) before a UO 2-rich solid solution with ThO 2 is precipitated. From equilibrium under a hydrogen atmosphere, which solubilizes the Pa 2O 5 phase as PaF 4, the equilibration quotient Q = X PaF 5 · P 1 2 H 2 X PaF 4 P HF was estimated as log Q = −3·92 + 5·16 ( 10 3 T ) . Entropies and formation free energies are estimated for PaF 4 and PaF 5 in solution and for the Pa 2O 5 phase.

Stability of BF 3OH − ion in molten and solid NaBF 4 and NaFNaBF 4 eutectics

I.R. measurements on solutions of NaBF 3OH in crystalline and molten NaBF 4 were employed to determine the stability of the BF 3OH − ion with respect to its unimolecular decomposition. Isotopic substitution experiments with D 2O, DF and D 2 have shown that the species BF 3OD − is formed in molten NaBF 4 either by reaction with BF 3OH −, with BFO 2O −, or with B 2F 6O 2− mediated by a transition metal oxide. These isotope reactions give encouragement to the view that tritium released during the operation of a molten salt breeder reactor might by trapped by reaction with BF 3OH − or with oxyfluoroborates present as impurities in the molten sodium fluoroborate coolant salt.

A Systematic Procedure for Reactor Control System Design

systematic design procedure was developed to aid in the reduction of trial and error in the design of reactor control systems. The method developed here uses an automatic optimization to help the designer find the best controller structure and the best controller settings to use in a control system based on conventional controllers (derivative, proportional, or integral). The method was demonstrated in a control system design for a molten salt breeder reactor. The application of the method gave excellent results at only a modest expense in computer time (5 min on the IBM-360 Model 91). The method appears suitable for routine application in reactor control system design.

Lanthanum Trifluoride as a Membrane in a Reference Electrode for Use in Certain Molten Fluorides

A reference electrode system which utilizes the unique properties of a single crystal lanthanum trifluoride membrane, namely, negligible solubility and virtually 100% current transport by fluoride ions, was successfully used both for electroanalytical measurements and for obtaining thermodynamic data in certain fluoride melts of interest to the Molten Salt Breeder Reactor.

The exchange of U 4 + and Th 4 + between molten LiF-BeF 2-ThF 4-UF 4 and (U-Th)O 2 solid solutions

The exchange of uranium and thorium between LiF-BeF 2-ThF 4 melts and (U-Th)O 2 solid solutions is relevant to the oxide chemistry of Molten Salt Breeder Reactor fuel, to the thermodynamics of UO 2-ThO 2 solid solutions, and to a two-phase “solvent extraction system” in reprocessing reactor fuels. Exchange equilibrium at 580–780 °C usually was obtained after 40–120 h of vigorous stirring of the two phases. The equilibrium quotient of mole fractions, Q = (X UO 2 X ThF 4 ) (X UF 4 X ThO 2 ) , was determined at various compositions of the oxide phase and of the fluoride phase and at different temperatures. The results obtained for 72-16-12 mol % LiF-BeF 2-ThF 4 were fitted by least squares to an equation containing only two parameters: log 10 Q = (2101 + 550 X UO 2 ) T. Other results indicated that Q did not vary over a fairly wide range of composition of the fluoride phase. The expression for Q gives for the oxide solid solution log 10 γ UO 2 = (275 T) X 2 ThO 2 and log 10 γ ThO 2 = (275 T) X 2 UO 2 , and for the heat of mixing ( ΔH m ) 0.5 = 315 cal/ mol.

Molten-Salt Reactors—History, Status, and Potential

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 fu...

Engineering Development of the MSBR Fuel Recycle

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 isolate 233Pa 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 to 233U. 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.

Foreword: The Status and Technology of Molten-Salt Reactors—A Review of Molten-Salt Reactor Work at the Oak Ridge National Laboratory

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.

The Design and Performance Features of a Single-Fluid Molten-Salt Breeder Reactor

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 diame...

Molten-Salt Reactor Chemistry

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-UF4 fuel 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 bred 233Pa and for removal of fission product poisons from a molten-salt breeder reactor is included.

Diffusion of titanium in modified hastelloy N

Diffusion coefficients of 44Ti in titanium-modified Hastelloy N have been determined over the temperature range of 800–1250 °C by standard procedures involving serial sectioning by latheing or grinding and counting by gamma spectroscopy. The data were fitted to the Arrhenius expression D = (15.3 ± 2.2) exp (−73 000 ± 3300/ RT) cm 2/ sec. Results were used to predict the maximum loss by diffusion of titanium from the alloy in a typical molten-salt breeder reactor at 700 °C. They are also compared with published diffusion data on similar systems.

Oak Ridge Uses U233 As Reactor Fuel

The world's first U233-fueled reactor (the Molten Salt Reactor Experiment) recently began operating at Oak Ridge National Laboratory. U233 could eventually be produced in quantity in molten-salt breeder reactors, in which nonfissionable thorium would be converted into fissionable U233.

Graphite behavior and its effects on MSBR performance

Graphite behavior under Molten-Salt Breeder Reactor (MSBR) conditions is reviewed and its influence on MSBR performance estimated. Based on the irradiation behavior of small-sized graphite specimens, a permissible reactor exposure for MSBR graphite is about 3 × 10 22 neutrons/cm 2 ( E > 50 keV). The stresses generated in the graphite due to differential growth and thermal gradients are relieved by radiation-induced creep, such that the maximum stress during reactor exposure is less than 1000 psi for reactor designs having a peak core power density of about 100 kW/liter and reactor exposures less than about 2 1 2 years. The corresponding power costs for single-fluid MSBR's would be about 3.1 mills/kWhr(e) based on a capital charge rate of 12% per year and an 80% load factor. Experimental data on graphite behavior also indicate that graphites with improved dimensional stability under irradiation can be developed, which would lead to improved reactor performance. The deposition of fission products on graphite does not appear to be large (10 to 35% of the “noble-metal” fission products based on MSRE experience); taking into account graphite replacement every two years, fission product deposition reduces the MSBR breeding ratio by about 0.002. Also, it appears that xenon poisoning can be kept at a 0.5% fraction poisoning level by using pyrolytic carbon as a pore impregnant which seals the surface of MSBR graphite and/or by efficient gas stripping of the fuel salt fluid by injection and removal of helium gas bubbles. It is concluded that good MSBR performance can be obtained by using graphite having combined properties presently demonstrated by small-size samples, and that development of MSBR graphite having such properties is feasible.

Fuel-Cycle Analysis of Molten-Salt Breeder Reactors

Thermal- and intermediate-energy molten-salt breeder reactors appear capable of fuel doubling times less than 20 years and fuel-cycle costs under 0.4 mill/kWh. These reactors are fueled with circul...

The Recovery of Protactinium and Uranium from Molten Fluoride Systems by Precipitation as Oxides

In the conceptual two-region molten-salt breeder reactor, fissionable U/ sup 233/ is recovered from the blanket as the decay product of Pa/sup 233/. Since equilibrium concentrations of Pa would result in appreciable parasitic neutron absorptions, the advantages of thermal breeding could be realized to a greater extent by removing both Pa/sup 233/ and U/sup 233/ from the blanket mixture. Methods for recovering these materials from molten-fluoride mixtures by precipitation as oxides are presented. Small-scale experiments clearly indicated that it is possible to remove protactinium from molten-fluoride solutions by a process that appears to be surface precipitation of protactinium on beryllium oxide or thorium oxide particles. Protactinium was removed from molten mixtures of LiF--BeF/sub 2/--ThF/sub 4/ (87-18-15 mole%) by the addition of l to 2% by weight of solid beryllium oxide or thorium oxide. The removal efficiency was high when the initial concentration of protactinium was either in the range 1 to 2 ppb or 50 to 75 ppm. Uranium was successfully removed from solution in molten fluorides by use of a similar procedure. Approximately 2000 ppm uranium was precipitated from molten LiF -BeF/sub 2/- ThF/sub 4/ (67-18-15 mole %) by the addition of 3% by weight of beryllium oxide. Comparablemore » results were also obtained using thorium oxide as the precipitant. (auth)« less