The current status of molten salt reactor development is discussed with reference to the experience from the Oak Ridge Molten Salt Reactor Experiment. Assessment of the future for this reactor system is reviewed with consideration of both advantages and disadvantages. Application of this concept to ADTT (accelerator driven transmutation technology) needs appears to be feasible by drawing on the MSRE experience. Key chemical considerations remain as: solubility, redox behavior, and chemical activity and their importance to ADTT planning is briefly explained. Priorities in the future development of molten salts for these applications are listed, with the foremost being the acceptance of the 2LiF-BeF{sub 2} solvent system. 8 refs, 2 figs

Dai, S.

Del Cul, G.D.

Gat, U.

Toth, L.M.

Williams, D.F.

UNT Digital Library

--, 7: .P 1 \J S 3y r@. k- 0ia, Review of ORNL’s MSR Technology and Status L. M. Toth, U. Gat, G. D. Del CUI, S. Dai, and D. F. Williams Oak Ridge National Laboratory, Oak Ridge, TI? 37831 BUG 1 6 199s Abstract. The current status of molten salt reactor development is discussed with reference to the experience obtained from the Oak Ridge Molten Salt Reactor Experiment. The assessment of the future for this reactor system is reviewed with both consideration of advantages and disadvantages. Application of this concept to ADTT needs appears to be feasible by drawing on the experience gained with the MSREi. Key chemical considerations remain as: solubility, redox behavior, and chemical activity and their importance to AD= planning is briefly explained. Priorities in the future development of molten salts for these applications are listed with the foremost being the acceptance of the 2LiF-BeF2 solvent system. The molten salt reactor experiment, MSRE, was operated from 1965-1969 and was shut down by draining the homogeneous fuel from the reactor circuit into two drain tanks located at a lower level in the MSRE facility. The purpose of this review is to identify the ORNL position with respect to MSR’s; review the pertinent MSR chemistry and the significant understanding gained since the MSRE operation era; and, finally, to relate these two elements to present accelerator driven transmutation technology, ADTT, interests. If one surveys the current attitude at ORNL with regard to MSR’s, he would observe that there is no official position. In parallel with the current national and DOE attitudes toward advanced nuclear programs, it is difficult to discern a definite position or proactive attitude toward the MSR status and technology. Rather, there is, at best, a “wait and see” attitude in spite of the fine history of development associated with this reactor concept and this attitude is consistent with the current DOE attitude toward all advanced reactor concepts. When viewed from outside, however, it is possible to become dismayed even though interaction with individuals or groups is still taking place. The most recent expression of a ORNL interest in MSR’s came in 1992 with documents written by F. J. Homan (then, director of reactor technology programs, ORNL) and U. Gat. While this expressed position is dated[l], it served to demonstrate the attitude toward MSR’s that still prevails with some individuals. At that time, Homan and Gat identified the MSR advantages as: (1) Simplicity: External cooling allows optimization of the core design to maximize efficiency. Molten salts have low vapor pressures at high temperatures, reducing k- the need for thick walled pipes and vessels. 8 N ofms tS UWI, This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or use- fulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any spe- cific commercial product, process. or service by trade name, trademark, manufac- turer, or otherwise does not necessarily constitute or imply its endorsement, recorn- mendotion. or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. . (2)  Ease of fuel handling: Fluid fuel can be moved with pumps and pipes eliminating the need for complex fuel handling machinery. (3) Simule fuel cvcle: No head end preparation for reprocessing, or complex waste management. The fuel is in fluid form, amenable to chemical manipulation. Molten salts can be purified using the very simple fluoride volatility process. (4) Simplified waste disposal: The waste which comes out of the reprocessing step contains little or no uranium or actinide elements. This simplifies waste management and results in better resource utilization than other reactor fuel cycles. ( 5 )  Versatility A single, basic design can be operated as modules, or in varying sizes; temperature control assures efficiency. One design can operate on any fuel. The MSR can burn actinides, burn plutonium from dismantled weapons, or breed. (6) Ease of shutdown and maintenance: Fuel can be drained for maintenance, reducing occupational exposure. (7) economic potential. Economy MSRs have been shown in various studies to have an excellent From today’s standpoint we might qualify waste disposal and economy realizing that these issues are now seldom simple or economical. Relative to other reactor systems, however, these points could be valid. On the negative side of the issue, Homan and Gat identified disadvantages and unresolved issues as: (1) (2)  (3) (4) (5) (6) (7) Lack of licensing experience. Possible reactivity fluctuations due to density variation or bubble formation. Necessity for a large external (to the core) fuel inventory, High radiation levels in primary system due to presence of fuel throughout. Reliability of components which contain and circulate the molten fuel. A not fully integrated (combined) chemistry. Lack of utility and industrial support. In addition to these listed issues, we might also add graphite swelling and undeveloped plutonium chemistry. As a result of this analysis, Homan and Gat identified the following position reflecting the O W L  attitude of the time: (1) The MSR has promise. (2) (3) The economic climate is not right to recommend the rebirth of MSR’s. A small program should be established to retain MSR experience lest retirement and/or death cause it to be lost. (4) MSRs might even be more attractive than LMFBR’s in today’s climate. Since the 1969 shutdown of the MSRE, several evaluations of MSR technology have occurred[2,3,4]. While many evaluations have been and will continue to be made, the primary requirements for the molten salt in a homogeneous fluid-fueled reactor still remain as expressed earlier.[5] The salt should have: (1) (2) (3) Low vapor pressure. (4) Low neutron cross section for the solvent components. Thermal stability of the salt components. Adequate solubility of fuel and fission product components. ( 5 )  (6) (7) Radiation stability. (8) It is readily seen from Table 2 or Ref. 5 that several of the fluoride salts satisfy the Adequate heat transfer and hydrodynamic properties. Chemical compatibility with container and moderator materials. Low fuel and processing costs. characteristic properties of thermal stability, manageable melting point and low neutron cross section. To achieve lower melting temperatures, two or more salts are combined to produce still lower melting mixtures. The most developed fluoride solvents consist of LiF and BeF, in 2:l mole ratio and melting at 452°C as shown in the phase diagram of Fig. 1 of Ref. 5. The 2LiF-BeF2 solvent has acceptable viscosity, low vapor pressure and good thermal stability for use as the solvent system in a molten salt reactor. In considering the application of these molten salts to a MSR situation, three important and interrelated chemical concepts must be controlled. These are solubility, redox chemistry, and chemical activity. Solubility as applied to molten salts involves more than what we ordinarily understand for aqueous systems. In the case of the pure fluoride materials, solubility is determined by phase diagrams of the mixtures. These temperature versus composition determinations give the temperature at which components might fall from solution (as determined by the liquidus line on the plot) and the identity of the salt phase which precipitates out. Because this information was of such vital importance to the control of the molten salt solution, much attention was given to the determination of phase diagrams in the early stages of the ORNL/MSR program. While much has been learned already about this phase diagram behavior, comparable information relating to PuF, is sorely lacking. For ADTT programs the determination of the Lif-BeF,-PuF, phase diagram - especially in the region of 2Lif-BeF2 with low concentrations of PuF,, is deemed as absolutely essential if PuF, fuels are to be used in the 2Lif-BeF2 solvent. In addition to concerns related to those of the pure materials, solubility has significance if impurities such as moisture react with the molten salts to produce metal oxides of much higher melting point and correspondingly lower solubility. For the case of the uranium he1 component, UF,, this reaction is written as: 2H,O + UF, * 4HF + UO, (1) where the very insoluble UO, could precipitate out. To prevent the loss of uranium from solution and possible accumulation of UO, precipitate in amounts large enough to pose criticality concerns, five mole percent of ZrF, was added to the solution to getter any oxide impurities through the reaction: 2H2O + ZrF, * 4HF + ZrO, which is more favorable to reaction with moisture than the former. Besides solubility concerns with respect to impurity reactions, one must prevent the liquid fuel component from reacting: (a) with the containment vessel and cause corrosion products plus structural weakening or (b) with the moderator graphite and leave behind insoluble fissionable materials. Reactions of the former such as: 2UF, + Cr (alloy component) * 2UF, + CrF, (3) and the latter: 4UF3 + 2C * UC, + 3UF, (4) were prevented by careful control of the solution redox chemistry which was accomplished by setting the UF,/UF3 ratio at approximately 100/1. Additions of metallic beryllium to the fuel salt to effect the reduction of UF, via: XJF, + Be" + XJF, + BeF, to achieve the desired UF,/UF, ratio. The routine adjustment of the redox chemistry was necessary to compensate for the net oxidizing effect of the fissioning uranium ; but since much of this chemistry has been described in earlier reports[5], the reader is referred there for these details. The significance of the redox control to ADTT systems is that in some cases where the fuel is PuF,, the Pu(III)/Pu(IV) redox couple is too oxidizing to present a satisfactorily redox “buffered” system. It is therefore proposed that redox control be accomplished by including an HF/H, mixture to the helium cover gas sparge which will not only set the redox potential, but will also serve as the redox indicator if the exit HFM, stream is analyzed relative to the inlet. For a 7-1 OMWT reactor system similar to the MSRE but run with a PuF,, it is calculated that as little as 1 cc/min HF would be required to compensate for the net reducing effect of plutonium fission. (See the previous presentation[6] describing the difference in redox changes brought about by PuF, fission in comparison to UF, fission.) The chemical activity of the molten fluoride solute components and the influence of molten salt solution composition on activity coefficients were actively considered during the MSRE era but the significance of solvent changes on these solute chemical equilibria was not strongly emphasized. The solvent system in the MSRE is, essentially, 2LiF-BeF2 which is “neutral” in a Lewis acidhase terminology with regard to the free fluoride ion concentration as determined by: 2LiF + BeF, BeFt- (6) Further emphasis of this important characteristic is valuable to ADTT because departure from the 2LiF-BeF2 composition has been contemplated in future concepts. One such system is NaF-ZrF, which is attractive because it does not present the health hazard of BeF,. However, the eutectic composition of NaF-ZrF, is highly acidic (ie., fluoride deficient) much like the LiF-BeF, eutectic composition the vapor pressure of the acidic component is significant. It was high enough in pre-MSRE days at Oak Ridge, that ZrF, was found to distill from the melt and condense on cooler surfaces in the containment system. Control of the ZrF, mass transport was considered too difficult to ensure, so the 2LiF-BeF2 solvent system was chosen. Figure 1 shows the effect of solvent changes on the equilibrium quotient for the hydrogen reduction of UF,: UF, + %H2 * UF, + HF (7) It can be readily seen that there is a significant change in the equilibrium quotient in going from the (Lewis) basic LiF-BeF2-ThF,(72-l6-12 mol %) to the acidic LiF-BeF,(48-52 mol %). Take for example the equilibrium quotients at 600°C for Lines B and D which are 5 .62~10-~ ORNL-OWG 76-127f3R -4 T ("C) 800 750 700 6 50 600 5 50 500 I I I I I I I SOLVENT mole 7. - A LiF-BeF2 -ThF, (72- \6 -12) -  8 LiF- BeF, (66-34) - D L i F - B e F 2  ( 4 8 - 5 2 )  - C LiF-BeF, ( 5 7 - 4 3 )  - EQUL I \  I I REACTION UF, + '/2 H z ~  UFJ+ HF [uF3] 'HF W H E R E  OR= ;; 0 MEAN OUTLET H F  ANALYSIS A MEAN INLET H F  ANALYSIS - AVERAGE L -8 I I 1 I I I I I I 0.90 0.98 I .06 I .i4 i .22 1.30 'ooo/7. ( O K  ) Fig. 1. Equilibrium quotients[7] for the H2 reduction ofUF,. Activity coefficient effect from solvent changes shown by shift in Q values, and 3 .47~10-~  atm”, respectively. This gives an equilibrium quotient ratio of 16.2. The practical significance of this difference is that one finds the UF, equilibrium concentration is considerably higher in the acidic (Line D) case. The change is due solely to the change in activity coefficients which is brought about by fluoride ion coordination equilibria as shown in Eq. 6 .  Figure 2 shows how the activity coefficients of the mono-di-tri and tetrapositive fluoride salts vary with LiF composition at 600°C. From these activity coefficients, one can calculate the expected change in the equilibrium quotient for changes in the solvent composition. However, these activity coefficient data are often limited and the prediction of their values at other temperatures is difficult because the coefficients are expected to vary considerably, depending on the complexation differences of the ions as a function of temperature. In ADTT applications these activity coefficient variations are important because changes in their values with solvent composition and temperature variations will have a profound effect on the equilibrium chemistry which controls redox potentia1, corrosion and fission product behavior. For example, in the more acidic NaF-ZrF, system, the chemistry of these processes will be much different from that experienced for the LiF-BeF, solvent and consequently should be carefully evaluated if it were to be selected. As has been stated previously[6], fuel processing to remove fission products needs further development. Krypton and xenon removal by gas sparge is an obvious necessity and easily achievable. Inclusion of a H,/HF mixture for redox control could provide a compatible combination to solve both objectives. Nevertheless, removal of other fission products must be considered if long-term operations are anticipated. It has been stressed several times already and bears repeating here --- on line processing to remove fission products should be considered only after having developed a satisfactory batch processing mode. Otherwise, coupling the complexities of a slip-stream processing operation to an operating accelerator- driven system would compromise the reliability of the overall system by adding a yet-proven process. Priorities that should be given in future molten salt development with regard to ADTT applications are, first, selection of 2LiF-BeF2 as the basic solvent system with lowest possible ZrF, oxide getter to minimize problems associated with reductive processing of the fuel salt and the associated reduction of ZrF,.. Next most important is the consideration of PuF, 2 0.5 0.2 0.5 0 6  0.7 0.C XLlf Fig, 2. Activity coefficients[8] in the LiF-BeF2-ThF, mixtures at 600°C. chemical behavior in this solvent system: PuF, solubility, the phase diagram behavior of the ternary PuF,-LiF-BeF, system in the region of low (<5 mole percent) PuF,, and the redox effects of the fission process. Analytical chemical support must be also included in the list of priorities as a key infrastructure element. Next in priority should be processing chemical development; first the establishment of the degree necessary with batch processing carrying the needs while continuous processes are developed over a longer time period. Finally, fission product chemistry behavior should be addressed with tellurium and noble metals as key points of concern. However, a full understanding of their chemical behavior could be deferred until the later stages of technology development since the other priorities hold a greater degree of preeminence. CONCLUSIONS Molten salt reactors are a proven concept and merit future development for ADTT applications. The molten fluoride chemistry for the 2LiF-BeF2 system is well established and can be applied with great confidence. However, other less understood solvent systems must be considered with caution lest subtle solvent effects such as those presented here cause severe problems with process operations. The chemistry of plutonium, as PuF,, needs further study and testing, especially in corrosion loop studies for redox control. Processing of the fuel salt needs further development with continuous on-line processing as the ultimate objective. Nevertheless, the future for systems utilizing this high-temperature fluid can be very promising based on the fine record established over the past several decades. ACKNOWLEDGMENTS This work has been sponsored by the Basic Energy Sciences, Division of Chemical Sciences, U. S .  Department of Energy under contract DE-AC05-960R22464 with Lockheed Martin Energy Research Corporation. REFERENCES Homan, F. J. and Gat, U., “Status Position on the Molten Salt Reactor,” June 1992 (unpublished). MacPherson, The Molten Salt Reactor Adventure, Nuclear Science and Engineering, 90, (1985). Rosenthal, M. W., Haubenreich, P. N., McCoy, H.E., and McNeese, L. E. , “Recent Progress in Molten Salt Reactor Development”, Atomic Energy Review, IX, 3,601, (1971). McNeese, L. E., et al, “Program Plan for Development of Molten Salt Breeder Reactors,” ORNL-50 18, December 1974. W. R. Grimes, “Chemical Research and Development for Molten Salt Breeder Reactors”, ORNL/TM- 1853, (1967). L. M. Toth, G. D. Del CUI, S. Dai, D. H. Metcalf, “Molten Fluoride Fuel Salt Chemistry”, First International Conference on ADTT and Applications, Las Vegas, Nevada, USA, July 25-29, 1994. Gilpatrick, L. O., and Toth, L. M., “The Hydrogen Reduction of UF, in Molten Fluoride Solutions,” J. Znorg. and Nuclear Chemistry, 39 (1 0), 18 17, (1 977). Baes, C. F., “Molten Salt Reactor Semiannual Progress Report”, ORNL-4548, Feb. 28, 1970, p. 152. 

Review of ORNL`s MSR technology and status

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Review of ORNL`s MSR technology and status

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