h - —— — — — — = — —— — _ — CENTRAL R ESEARCH LI DOCUMENT COLLECTIBC?P? i OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION m NUCLEAR DIVISION for the U.S. ATOMIC ENERGY COMMISSION [ g ) - | 445k 05 1] MATERIALS FOR MOLTEN-SALT REACTORS | . | H. E. McCoy W. H. Cook R. E. Gehlbach J. R. Weir, Jr. C. R. Kennedy C. E. Sessions R. L. Beatty A. P. Litman J. W. Koger | o [s | NOTICE This document contains information of a preliminary nature and was preparec primarily for internal use at the Oak Ridge Mational Laboratory. It is subject to revision or correction and therefore does not represent a final report. 1 — LEGAL NOTICE This repart wos prepared os an account of Government sponsared work. Neither the United States, nor the Commission, nor any persen octing on behalf of the Commission: A. Mokes ony waorranty or representation, expressed or implied, with respect to the occurocy, completeness, or usefulness of the informotion contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report moy nel infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, opparatus, method, or process disclosed in this report. As used in the above, ‘person acting on behalf of the Commission" includes any employee or contractor of the Commission, or employee of such contracter, to the extent that such employee or contractor of the Commission, or emplovee of such contractor prepares, disseminates, or pravides access fo, any infermation pursuant ta his employment or contract with the Commission, or his employment with such contractor. ORNL-TM-2511 Contract No. W-7405-eng-26 METALS AND CERAMICS DIVISION MATERTALS FOR MOLTEN-SALT REACTORS H. E. McCoy W. H. Cook R. E. Gehlbach J. R. Weir, Jr. C. R. Kennedy C. E. Sessions R. L. Beatty A. P. Litman J. W. Koger Paper submitted to the Journal of Nuclear Applications MAY 1969 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the - U.S. ATOMIC ENERGY COMMISSION ' - - | LOCKHEED MARTIN ENERGY RESEARCH LIBRARIES t | 3 445k 0513k7kL 9 iii CONTENTS Abstract . . . . . . . . .. 00 . Introduction . Experience With The MSRE Development of A Modified Hastelloy N With Improved Resistance to Irradiation Damage . . . . . . . . Irradiation Damage in Graphite . . . . . SUMMATY v v ¢ & o o o o o o o o o o o o o o o « o« o o« References . . 14 35 36 MATERIALS FOR MOLTEN-SALT REACTCRS H. .E. McCoy W. H. Cook R. E. Gehlbach J. R. Weir, Jr. C. R. Kennedy C. E. Sessions R. L. Beatty A. P. Litman J. W. Koger ABSTRACT Operating experience with the Molten-Salt Reactor Experiment (MSRE) has demonstrated the excellent compat- ibility 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, (1) Hastelloy N with improved resistance to embrittlement by thermal neutrons, (2) graphite with better dimensional stability in a fast neutron flux, (3) graphite that is sealed to obtain a surface permeability of less than 10”8 cm?/sec, and (4) a secondary coolant that is inexpensive and has a melting point of about 400°C. A brief description is given of the materials work in progress to satisfy each of these requirements. Our studies presently indicate that significant improvements can likely be accomplished in each area. INTRODUCTION Our presenf concept of & molten-salt breeder reactor (described in detail by Bettis et 5;.) utilizes graphite as moderator and reflector, Hastelloy N for the containment vessel and other metallic parts of the system, and a liquid fluoride salt containing LiF, BeF,, UF,, and ThF, as the fertile-fissile medium. The‘fertile-fiSSile salt will leave the reactor vessel at a temperature of about 700°C and energy will be trans- ferred to a coolant salt which in turn is used to produce supercritical steam. Our experience with the Molten-Salt Reactor Experiment (MSRE) has demonstrated the basic compatibility of the graphite-Hastelloy N-fluoride salt (LiF-BeF,-ZrF,-UF,) system at 650°C. However, a breeder reactor will impose more stringent material requirements; namely, (1) the design life of the basic plant of a breeder is 30 years at a maximum operating temperature of 700°C, (2) the power density will be higher in a breeder and will require the core graphite to sustain higher damaging neutron flux and fluence, and (3) neutron economy is of utmost importance in the breeder and the retention of fission products {particularly *°Xe) by the core graphite must be minimized. ZEach of these factors requires a specific improvement in the behavior of materials. We have found that the mechanical properties of Hastelloy N deteriorate as a result of thermal neutron exposure and must find a method of improving the mechanical properties of this material to ensure the desired 30-year plant life. Similarly, we have found that graphite 1s damaged by irradiation. Although we can replace the core graphite, we find that the allowable fast neutron fluence for the graphite has an important infliuence on the economics of our reactor. Thus, we have undertaken a program to learn more about irradiation damage in graphite and to develop graphites with improved resistance to damage. A big factor in neutron economy is reducing the quantity of 135y%¢ that resides in the core. We will remove this gas by continuously sparging the system with helium bubbles, but the transfer by this method probably will not be rapid enough to prevent excessive quantities of 135Ye from being absorbed by the graphite. This can be prevented by reducing the surface diffusivity to < 1078 cm®/sec, and we feel that this is best accomplished by carbon impregnation by internal decomposi- tion of a hydrocarbon. We are also searching for a new secondary coolant that will allow us greater latitude in operating temperature. Sodium fluoroborate has reasonable physical properties for this application, and we'ére evaluating the compatibility of Hastelloy N with this salt. We shall describe our work in each of these areas in some detail. EXPERIENCE WITH THE MSRE Other papers in this series have elaborated on the information ga;ned from the MSRE regarding operating experience, physics, chemistry, and fission-product behavior. Additionally, valuable information has been gained about the materials involved.(l_B) We haye'a surveillance facility exposed to the salt in the core of the reactor and one outside the reactér vessel, where the environment is nitrogen plus about 2% 0,. Hastelloy N tensile rods and samples of the grade CGB graphite* used in the core of the MSRE are exposed in the core facility. The components are assembled so that portions can be removed in a hot cell, new sémples added, and returned to the reactor. We have removed samples after llOO,I44OO, and 9100 hr of full-power (8 Mw) operation at 650°C. As shown in Fig. 1, the physical condition of the graphite and metal samples was excellent; identification numbers and machining marks were cledrly visible. The peak fast fluence received by the graphite has been 4.8 x 1029 neutrons/cm® (> 50 keV) and the dimensional changes are less than 0.1%. Pieces of graphite from the MSRE have been sectioned and most of the fission products were found to be located on the surface and within 10 mils below the surface. *Trade name of Union Carbide Corporation for the needle-coke graphite used in the MSRE. W R-42962 et Fig. 1. Graphite and Hastelloy N Surveillance Assembly Removed from the Core of the MSRE After 72,400 Mvhr of Operation. Exposed to flowing salt for 15,300 hr at 650°C. However, a few of the fission products have gaseous precursors and penetrated the graphite to greater depths. The microstructure of the Hastelloy N near the surface was modified to a depth of about 0.001 in., but we found a similar modification.in samples exposed to static non- fissioning salt for an equivalent time. We have not positively identified the near-surface modification, but find its presenée of no consequence. The very small changes in the amounts of chromium and iron in the fuel salt also indicate very low corrosion rates and support our metallographic observations. Hastelloy N samples were removed from the surveillance facility outside the reactor vessel after 4400 and 9100 hr of full-power operation. This enviromment is oxidizing, and we have found that an oxide film about 0.002 in. thick was formed on the surface after the longer exposure. There was no evidence of nitriding, and the mechanical properties of these samples were not affected adversely by the presence of the thin oxide film. Thus, our experience with the MSRE has proven in service the excel- lent compatibility of the Hastelloy N-graphite-fluoride salt system. DEVELOPMENT OF A MODIFIED HASTELLOY N WITH IMPROVED RESISTANCE TO TIRRADIATION DAMAGE Since the MSRE was constructed, we have found that Hastelloy N, as well as most otfier iron- and nickel-base alloys, is subject to a type of high-temperature irradiation damage that reduces the creep-rupture strength and the fracture strain.(4_9) This effect is characterized in Figs. 2 and 3 for a test temperature of 650°C. The rupture lives for irradiated and unirradiated materials differ most at high stress levels ORNL-DWG 68-4200 70 N HEAT NOS. N o -5065 \ A-5067 ' o - 5085 60 . N ' "~ PRETEST ANNEAL-IHR AT 1177°C \\ SOLID POINTS-IRRADIATED <150°C 50 N OPEN POINTS-IRRADIATED 500— . 650° G * AN o N§f . \ | ;\\ a8 & < < \\ ® ,/ __.-—'// {’ T"h- ::_,4— 0 .00 0t A . | 10 100 MINIMUM CREEP RATE, %/HR Fig. 5. Variation of Fracture Strain with Strain Rate for Several Hastelloy N Type Alloys. Samples irradiated to a fluence of about 5 x 10?° neutrons/cm?® prior to testing. 14 TRRADIATION DAMAGE IN GRAPHITE Neutron irradiation alters the physical properties of graphite, but our major concerns arise because of the dimensional changes that 15,16) occur.( These dimensional changes are illustrated in Fig. € where the data of Henson et 2&,(17) are presented for an isotropic graphite. With increasing fluence the graphite first contracts and then begins to expand at a very high rate. Several potential problems arise as a result of these dimensional changes. First, the initial contraction will change the volume occupied by fuel salt and change the nuclear characteristics of the reactor. These dimensional changes seem small enough for most isotropic graphites that the nuclear effects may be accommodated by design. A second problem is stress generation due to flux gradients across a piece of graphite. Graphite creeps under irra- diation(lg) and we have shown that this creep is large enough to reduce the stress intensities to quite acceptable values. The third and most serious problem is that the rapid growth rate represents a rapid decreasé in density with potential crack and void formation. At some fluence this will cause the mechanical properties to deteriorate and the perme- ability to salt and fission products to increase. We feel that the properties will be acceptable, at least until the material returns to its original volume, and have defined this fluence as the lifetime. A fourth problem is that the dimensional changes are dependent on temper- ature and the curve in Fig. 6 is shifted up and to the left for increasing temperature. Thus, stresses develop in a part having a temperature gradient since segments of the part are seeking different dimensions. Again, this stress is relieved by the irradiation induced ORNL-DWG 67 —10504A 10 o 550~-600° C / VOLUMETRIC CHANGE (%) O . v e l O, \\ _/_// 400-440°C —— -10 0 10 20 30 40 50 (x402) NEUTRON FLUENCE (neutrons/cm?) (£ >50 keV) Fig. 6. Volume Change in Isotropic Graphite Dounreay Fast Reactor Irradiations. GT 16 creep in graphite geometries of interest to us. Therefore, we consider the onset of rapid growth to be the primary problem and the initial dimensional changes of secondary importance. We anticipate graphite temperatures between 550 and 750°C and would like to operate with a fast flux (> 50 keV) of about 1 X 10'° neutrons cm™? sec™!. Data in Fig. 6 show that this flux will cause this particular graphite to expand rapidly after a fluence of approximately 3 X 10°° neutrons/cm? is reached (about 1 year of operation). We can reduce the flux by decreasing the power density, but this is done only with increases in the fuel inventory and doubling time. Hence, it is quite desirable that we use graphites with better resistance to irra- diation damage than the graphite shown in Fig. 6. We examined the data available on current reactor graphites irradiated to high fluences and found that the results described a fairly consistent picture. The fluences required for graphite to reach its minimum volume were strongly temperature dependent (decreased with increasing temperature), but were not appreciably different for any of the graphites studied to date. Although this observation is discouraging, we have pressed the problem further and have found that better graphites already exist and that others can probably be developed with only small changes in present mate- rials and processing. ILet us look briefly at a simple description of the origin of the dimensional changes and then return to our specific observations. Graphite, after being well graphitized at temperatures above 2000°C, has a hexagonal close-packed crystal structure consisting of close-packed layers (basal planes) of carbon atoms with very strong covalent bonds 17 within the basal planes (a-direction) and very weak van der Waals' forces between atoms in adjacent basal planes (c-direction). This anisotropy in atomic density and bond strength is reflected by very anisotropic properties. The changes that take place in a single crystal of graphite during irradiation are shown schematically in Fig. 7. A neutron having an energy above about 0.18 eV can displace a carbon atom from a close-packed basal plane with a reasonable probability of creating a vacancy in a basal plane and an interstitial carbon atom between the basal planes. Repetition of this process. and diffusion at elevated temperatures can result in the formation of defect clusters, specifically partial planes of atoms between the basal planes and vacancy clusters within the original planes. This leads to an expansion perpendicular to the basal planes (c-direction) and a contraction within the layer planes (a- direction) — see Fig. 7(b). This new configuration leads to a slight increase in volume [Fig. 7(e)]. Polycrystalline graphites are not initially of theoretical density. The voids present in the material are a result of shrinkage of the binder during grephitization and from separation or fracture of layer planes during cooling from the graphitization temperature. Initially,vduring irradiation the porosity within the material tends to be filled as the crystallites expand in the c-direction. This produces a densification illustrated by the bottom curve in Fig. V(d). As the porosity fills, the shrinkage saturates and the dimensional behavior begins to be dominated by the volume expansion due to the growth of the crystallites in the c¢-~direction. Thus, a minimum volume is obtained as shown in 18 ORNL —DWG 68— 42011A PERFECT SINGLE CRYSTAL DEFECT - * PLANES g — AL av Lo ® N, BASAL . PLANES (@) (b) (¢) POLYCRYS TAL DEFECT PLANES + + OBSERVED AV | AV Vo T, 2 PREDICTED - DENSIFICATION - (d) (e) Fig. 7. Graphite Dimensional Changes Due to Irradiation. 19 Fig. 7(e). During the subsequent expansion, the material either remains internally contiguous, in which case the volume change rate of the poly- crystalline material should be similar to the small rate of expansion exhibited by the crystallites themselves, or fractures internally due to the stresses generated between the crystallites of differing orienta- tion (causing a higher rate of growth to occur). Observations to date indicate that most graphites increase in volume at a faster rate at high fluences than expec£ed if the material remained internally contiguous. One further consideration helps ué to understand why the unpredicted rapid growth takes place. A schematic representation of several coke particles and binder after graphitizatién is shéwn in Fig. 8 (ref. 19). Each coke particle consists of several crystals with a very high degree of preferred orientation. Although we make a large piece of graphite in which the coke particles are arranged randomly, there are still inter- faces between particles of widely different orientations. As each particle changes dimensions, these interfaces must be strong and able to shear large amounts without fracturing. The observation that graph- ites undergo large dimensional changes at high fluences indicates that these interfaces or boundaries are fracturing. As indicated by the sketch in Fig. &, these boundaries are made up largely of the graphitized binder materials. Thus, the properties of these boundaries are influenced largely by the nature of the binder material and its interaction with the coke particles. We have set about making graphites with known filler and binder materials, but our work in this area has not progressed very far. We are also studying the properties of several commercial graphites that we feel may be potentially useful for MSBR applications and others that GRAPHITE Petroleum coke Pitch coke crystallites crystallites Petroleum / coke particle Fig. 8. Proposed Arrangement of Crystallites in Graphitized Stock. (Reprinted from High-Temperature Materials and Technology, ed. by I. E. Campbell and E. M. Sherwood,, Wiley, New York, 1967, p. 195). Qc 21 .should give us some basic information about irradiation damage in graph- ite. Our graphite irradiations have been done at 705 + 10°C in the High Flux Isotope Reactor (HFIR) Whefe the peak flux (> 50 keV) is 1 X 101° neutrons cm™? see™?. Thus, samples can be irradiated to fluences of 1 X 1072 neutrons/cm2 in about 4 months. A summary of our results to date is shown in Fig. 9. Several mate- rials show a significant deviation from the "typical" behavior illustrated in Fig. 6. The POCO* graphites show excellent resistance to irradiation with very small positive dimensional changes out to fluences of 2.5 X 10%? neutrons/cm®. Thus, these data give us confidence that the "typical" behavior of graphite can be improved markedly, but our results have not been extended to fluences high enough to determine the exact magnitude of this improvement. SEALING GRAPHITE Entry of fuel éalt into graphite can be prevented by keeping the entrance_diameter of the accessible porosity smaller than 1 p. Although this does require some extra care during processing, it can be accom- plished routinely on large shapes. In fact, the grade CGB graphite obtained 5 years ago for the MSRE satisfies this requirement.(2°) How- ever, the graphite structure must be much more restrictive to prevent gaseous fission products, particularly '3°Xe, from diffusing into the graphite. We presently plan to strip '?°Xe from the fuel salt by purging with helium. Helium bubbles will be injected and later removed in a gas-liquid separator. The efficiency of this purging depends very *POCC Graphite, Inc., Garland, Texas. 4OOIn(eV +1) 0 Fig. 9. 22 ORNL-DWG68-10063R3 J ) Y-12-3 BY-12 / /] / — POCO GRADES T / o N \_/ N N\ UK-ISO\\ NAT. ATJ-SG oA H-315A I 7 %7‘" / // 7 (BNWL) AGOT / i425>\ ATJ—S% /50kev) 25 30 (X102 Volume Changes of Graphite Irradiated at 705°C. 23 heavily on the size of bubbles that can be injected and cifculated and the mass transfer of '?°Xe from the salt to the helium bubbles. Both of these factors are presently uncertain, and we must anticipate that large quantities of '°°Xe will be available to the graphite surfaces and that excessive (> 0.5%) retention of !?°Xe will result if this gas can enter the graphite surface at a high rate. Our present calculétions show that the accessibility of ?°Xe to the graphite surfaces will be 'impeded by a laminar salt film and that the graphite offers an additional resistance to gas flow only when its surface diffusivity to 135%e is less than 1078 cmz/sec._ The best grades of commercial graphites presently available have bulk diffusivities in the range of 10"1 o 1074 cm2/sec, and we feel that it is unreasonable to expect that techniques can be developed for making massive shapes with such a restrictive structure. The techniques used for reducing the porosity of graphite involve multiple impregnations of the material with liquid hydrocarbons and then firing to graphitize this material. As the bulk diffusivity decreases, it becomes progres- sively more difficult for the gases released by the decomposing impreg- nants to diffuse out of the material and the times required to reach the graphitizing temperature become excessive. Thus, we presently feel that it is more reasonable to reduce the surface diffusivity by &a post- fabfication surface-sealing process involving gaseous impregnation. Since the pyrolytic carbon that would be deposited and the graphite sub- strate will change dimensions differently under irradiation, it is imperative that the pyrolytic carbon be linked with the substrate struec- ture and not deposited as a surface layer that can be sheared easily. 24 The task that we have in sealing the graphite is illustrated by the photomicrograph in Fig. 10. We must adjust our processing parameters so that the volds are filled internally in preference to closing the voids near the surface and forming a coating. This can be accomplished by using a flowing stream of hydrocarbon at low partial pressure and temperature appropriate to maintain very low deposition kinetics, but this requires long processing times. We have used a different method to accomplish penetration which involves pulsing the sample environment between a rich hydrocarbon environment and vacuum. The vacuum cycle removes the reaction products (primarily hydrogen) and allows more hydro- carbon gas to enter the void. Specifically, we have used 1,3 butadiene at 20 psig, deposition temperatures of 800 to 1000°C, and cycle times of about 1 min for the vacuum and a fraction of a second for the hydro- carbon. Butadiene was chosen because it is a gas at room temperature and because of its high carbon yield per molecule. The temperature range is restricted to 800 to 1000°C because higher temperatures result in a surface coating not penetrating the pore structure and lower tem- peratures yield intolerably low deposition rates. The lengths of the vacuum and pressure periods are very important because they not only influence the processing rate, but also the depth of penetration of the impregnant. The time required for the process will be very important in determining the cost. We have worked with two commercial graphites — viz, AXF made by POCO and ATJ-SG made by UCC. These materials had widely different accessible pore spectra; nearly all the pores in the AXF material were less than 0.8 p in diameter while the ATJ-SG grade had appreciable pores 25 Fig. 10. Photomicrograph of the Edge of Graphite Showing a Pore that has been Partially Coated and then Sealed over with Pyrocarbon. 26 in all size ranges up to 17 p. Thus, the sealing characteristics of the two materials were widely different. The results of some of our parameter studies are shown in Fig. 11 where we have varied the vacuum- hydrocarbon cycle times at 850°C. The initial slopes are proportional to the surface area being coated and the slope is much steeper for the AXF graphite than for the ATJ-SG material. The sharp break in the curves for the AXF graphite indicates that the pores have been filled or closed off and that the surface area being coated is reduced. The sharpness of this break attests to the uniform pore size of the AXF graphite. The horizontal portions of the curves represent essentially surface coating, and the data suggest that some finite amount of surface coating is necessary to attain the MSBR permeability specificdtion of less than 1078 cm®/sec for 135%e at 700°C. This is approximately equiva- lent to a helium permeability of 108 cmz/sec at ambient temperature. We have used the latter criterion in our studies (denoted on these curves by a "v" mark). The ATJ-SG graphite was not sealed to the desired level under the conditions shown in Fig. 11 and the slope changes very gradually due to the wide variation in the pore sizes. The data indicate that the processing time could be reduced by shortening the length of the vacuum cycle. Another interesting feature of the process for the AXF graphite was that the final total weight of carbon deposited was increased by shortening the vacuum pulse. This indicates that the depth of penetration of carbon into the material was increased. Thus, shortening the vacuum pulse accelerated the process and improved the product, both very desirable characteristics. 27 0.10 0.08 0.06 0.04 MATERIAL: POCO AXF GRAPHITE o O N ORNL-DWG 68-12030 VACUUM (sec) HYDROCARBON (sec) 60 60 30 30 15 10x10°8cm?2 /sec (HELIUM) <1079 m2/sec (HELIUM) TOTAL WEIGHT GAINED (g) o 0 O T _4| T T T T T T M5X1O mm Hg(AIR) 2.5%x102mm | L oHg(AIR) o —425x1072mm = 48x1072 mm —*" |Hg (AIR) Hg (AIR) 008 A .,/ " '—-_J-_-—-T ( -t o] /‘//x -2 A o At 1 75x10" “mm > o L= e Hg (AIR) /o/ / /1,__4/ 006 -.,“ 1/ /“/ V p 7 > L~ o ./.,j/ /‘/‘/ '/ ,/ A | 0.04 ol yed ;‘ ‘/n / . // _ Ve MATERIAL: ATJ-SG GRAPHITE 002 AN /o Y/ /7 0 0 4 8 12 16 20 24 28 32 36 40 46 TOTAL PROCESSING TIME (hr) Fig. 11. Impregnation Rate of Graphite Using 1,3 Butadiene at 850° C. 28 Our studies are not yet extensive enough to optimize the deposition conditions but are sufficient to make us optimistic about being able to reduce the surface diffusivity of graphite to the desired level. The remaining question of prime importance is the integrity of the seal after exposure to high neutron fluences. CORROSION IN FLUORIDE SALT SYSTEMS (21-29) ang our experience with the Two decades of corrosion testing MSRE(2’3) have demonstrated the excellent compatibility of Hastelloy N and graphite with fluoride salts containing LiF, BelF,, ThF,, and UF,. Our fertile~fissile salt will contain these éame fluorides, so only proof-testing will be required for the primary reactor circuit. How- ever, we desire a lower melting coolant salt in the secondary coolant circuit than the LiF-BeF, salt presently used in the MSRE and have chosen a sodium fluoroborate salt (NaBF,—8 mole % NaF) for further study. This salt is inexpensive (< $0.50/1b) and has a low melting point of 380°C. A significant characteristic of this salt is that it has an appreciable equilibrium overpressure of BF; gas (e.g., 180 mm at 600°C). Much of our present corrosion work is concerned with the compati- bility of Hastelloy N with sodium fluoroborate. Some earlier thermal convection loop studies involving a relatively impure salt of composition NaBF;~4 mole % NaF—6 mole % KBF, showed that a Croloy 9M loop plugged after 1440 hr at a maximum temperature of 607°C and a tempgrature difference of 145°C, and that a Hastelloy N loop was partially plugged after 8765 hr of operation under the same temperature conditions.(3°) The plug in the Croloy loop was comprised primarily of pure iron 29 crystals and the partial plug in the Hasteiloy N loop was made up of a compact mass of green single crystals of Na;CrF,. The salt charge from the Hastelloy N loop contained large amounts of Cr, Fe, Ni, and Mo, all major alloying elements in Hastelloy N. - In our more recent tests we have used a fluoroborate salt of compo- sition NaBF,—8 mole % NaF of a higher purity than the first salt that we worked with.(31’32) We are using a modified thermal convection loop from which we can remove salt samples for chemical analysis and metal samples for weighing without interrupting operation of the loop. The weight changes for the hottest and coldest samples are shown in Fig. 12 for the two loops presently in operation. The loops are constructed of identical materials, but the removable samples in one loop (NCL-13) are standard Hastelloy N and those in the other loop (NCIL-14) are a modified Hastelloy N containing 0.5% Ti and only 0.1% Fe (standard Hastelloy N contains 4% Fe). As shown in Fig. 12, the weight changes of the modified Hastelloy N are smaller, and this is later shown to be due primarily to the lower iron content of the modified material. The rate of weight change was steady except for a small perturbation after 1500 hr of opera- tion and a large variation after 4200 hr of operation. These times corresponded to times when moist air inadvertently came in contact with the salt. The changes in chemistry shown in Fig. 13 also reflect the admission of air at these times since the oxygen and water levels in the salt increased. The iron and chromium concentrations have continued to increase at a rate pfoportional to (time)%, indicating that the process is controlled by diffusion in the metal. The nickel and molybdenum con- centrations in the salt have remained very low except for times when 30 ORNL -DWG68- 12034 6 4 COLDEST SPECIMENS —— A 460°C / | o A 0/ A\ e 2 A,A——-'A"‘ S — W/ o—T e WEIGHT CHANGE (mg/cm?) t N ‘>/ 7 -6 HOTTEST SPECIMENS - 607°C -8 O—TITANIUM MODIFIED HASTELLOY N- NCL-14 -{0 —A—STANDARD HASTELLOY N- NCL-{3 \o -2 , 0o { 2 3 4 5 2 (x103) TIME ( hr) Fig. 12. Comparison of the Weight Changes of Hastelloy N Specimens Inserted in NaBF,-NaF {92—8 mole %) Thermal Convection Loops. 31 ORNL-DWG 68—-11768R 3000 ' ‘ 2000 >0 CONCENTRATION {(ppm) :‘:. 0 » 1000 F———j NI T ~ — . . - . ® ® 0 400 O 350 A- " A/ 300 e b A * E . ////// S 250 = 9 H g 200 o - [ ] = L g 150 * 8 s 100 7' //wc \QT,M N 50 HY C// )——0——0 O - 0 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 TIME (hr) Fig. 13. Variation of Impurities with Time in NaBF,-NaF (928 mole %) Thermal Convection Loop NCL-14. 32 moist air was inadvertently contacted with the salt. These results show qualitatively that the corrosion rates increased when the oxygen and water levels increased. Capsule tests in which sodium fluoroborate containing 1400 ppm O, and 400 ppm H,O was contacted with Hastelloy N for 6800 hr at 607°C exhibited very low corrosion rates. Future work will be directed toward defining the oxygen and water levels that result in acceptable corrosion rates. The information that we obtained on the changes in salt composition and the weight ;hanges of ten samples located at various points (and tem- peratures) around thé loop enabled us to attempt a mass balance for the system. The weight of metal lost must equal the weight of metal deposited plus the weight of metal in the salt — i.e., Awloss - Awdeposited * AwSalt ’ (1) We construct a weight change versus temperature profile based on the- removable samples and assume that each segment of the loop wall follows this same curve. This procedure results in mass balances, Eq. (1), that close within 10%. Diffusion theory can be used for further analysis. As mentioned earlier, chemical analyses (Fig. 13) indicate that the iron and chromium concentrations in the salt are increasing, so it is assumed that the salt selectively removes these elements from the alloy. The modified Hastelloy N is relatively free of iron, so the weight loss of this sample should be due primarily to the removal of chromium. The quantity of mate- rial removed by diffusion under conditions where the surface concentration of the diffusing element is zero is given by: where : 2 material removed, g/cm 5 & Cp = bulk concentration of diffusing species, g/cm3, o I diffusivity, cm®/sec ct Il time, sec. Using the data of Grimes et a_l.(33) for diffusion of chromium in Hastelloy N, we found by this analysis that the quantity of chromium removed by diffusion cannot account for the total weight lost. This discrepancy can be accounted for by short-circuit diffusion mechanisms enhancing the rate of chromium removal at these low temper- atures or by impurities (likely HF formed by water ingestion) that lead to some general attack of the metal that is nof diffusion controlled. Two observations argue against the latter possibility. First, we have not been able to see by microprobe analysis any transfer of nickel or molybdenum to the colder surfaces of the loop except during a brief period after 4200 hr of operation in which we knew that large amounts of impurities. were present. A second and more convincing argument is based on the relative behavior of the standard and modified Hastelloy N during the first 4000 hr of operation. Rewriting of Eq. (2) in terms of a reaction rate constant, K, instead of the diffusion coefficient yields M=C, JKt . (3) We have already indicated that this equation predicts the material trans- ported by diffusion to be too low for the modified alloy when K equals D, but let K take on a value so that the predicted and observed weight losses 34 for a given time of operation agree with Cg equal to 7% Cr. Now consider the standard alloy in which Cy corresponds to 7% Cr + 4% Fe. We find that the same K chosen for the modified alloy predicts the observed weight change for the standard alloy. Thus, the difference in the weight losses between the modified and sténdard alloy is due principally to the iron content. Had much general corrosion occurred, this adjustment'in Co should not have worked. In fact, we found this analytical procedure to be entirely unsatisfactory for the short time period after 4200 hr when the water and oxygen levels were high and nickel and molybdenum were being removed (Figs. 12 and 13). A further possible role of impurities is to provide the oxidizing potential necessary to keep'the surface con- centration of iron and chromium at zero. Thus, even though the process remains diffusion controlled, the rate can be increased by impurities. Although some of the curves in Figs. 12 and 13 have quite large slopes, the corrosion rates are not very high. Using the rather inac- curate method of converting the weight losses to corrosion depths indi- cates that the average rate during 8000 hr of operation has been 0.7 mil/yr. The rate has decreased to about 0.3 mil/yr for long time periods in which operation was not disturbed. In scaled-up MSBR systems we probably will use a cold trapping technique to remove some of the corrosion products so that their solubilities are not exceeded. We presently feel that the sodium fluoroborate salt will provide a satis- factory and economical secondary coolant for molten salt reactors. 35 SUMMARY OQur experience with the MSRE has proven the basic compatibility of the graphite-Hastelloy N-fluoride salt system at elevated temperatures. However, a molten-salt breeder reactor will impose more stringent operating conditions, and we need to make some improvements in the graphite and Hastelloy N for this system. The mechanical properties of Hastelloy N deteriorate under thermal neutron irradiation, and we have found that the addition of titanium in combination with strong carbide formers such as niobium and hafnium makes the alloy more resistant to this type of irradiation damage. Graphite undergoes dimensional changes due to exposure to fast neutrons, and‘the possible loss of structural integrity due to these dimensional changes presently limits the lifetime of the core graphite. Although we can replace the core graphite as often as neceséary, these replacements influence the economics of our reactor, and we have embarked on a program to find a better graphite. Our studies to date indicate that graphites can be developed that have better resistance to irradiation damage than conventional nuclear graphites. We plan to seal the graphite used in the core with pyrocarbon to reduce the amount of '?°Xe that is absorbed. Techniques have been developed for this sealing, and studies are in progress to determine whether the low permeability is retained after irradiation. Our corrosion studies are currently concentrated on evaluating the compatibility of Hastelloy N with a potential coolant salt, sodium fluoroborate. Cur studies indi- cate that the corrosion rate is acceptable as long as the éalt does not contain large amounts of impurities, such as HF and H;O. 10. 11. 36 REFERENCES W. H. Cook, Molten-Salt Reactor Program Semiann. Progr. Rept. August 31, 1965, ORNL-3872, Oak Ridge National Laboratory, pp. 87-92. H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimens — First Group, ORNL-TM-1997, Oak Ridge National Laboratory (November 1967). H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastelloy N Surveillance Specimens — Second Group, ORNL-TM-2359, Oak Ridge National Laboratory, in press. D. R. Harries, J. Brit. Nucl. Energy Soc. 5, 74 (1966). W. R. Martin and J. R. Weir, pp. 251-267 in Flow and Fracture of Metals and Alloys in Nuclear Environments Spec. Tech. Publ. 380, American Society for Testing and Materials, Philadelphia, 1965. J. T. Venard and J. R. Weir, p. 269 in Flow and Fracture of Metals and Alloys in Nuclear Enviromments Spec. Tech. Publ. 380, American Society for Testing and Materials, Philadelphia, 1965. W. R. Martin and J. R. Weir, Nucl. Appl. 1(2), 160-167 (1965). W. R.- Martin and J. R. Weir, Nucl. Appl. 3, 167 (1967). H. E. McCoy and J. R. Weir, Nucl. Appl. 4, 96 (1968). P.C.L. Pfeil and D. R. Harries, p. 202 in Flow and Fracture of Metals and Alloys in Nuclear Environments Spec. Tech. Publ. 380, American Society for Testing and Materials, Philadelphia, 1965. P.C.L. Pfeil, P. J. Barton, and D. R. Arkell, Trans. Am. Nucl. Soc. 8, 120 (1965). 12. 13. 14, 15. 16. 17. 18. 19. 20. 21. 22. 37 P.R.B. Higgins and A. C. Roberts, Nature 206, 1249 (1965). H. E. McCoy Jr., and J. R. Weir, Jr., Materials Development for Molten-Salt Breeder Reactors, ORNL-TM-1854, Oak Ridge National Laboratory (June 1967). H. E. McCoy, Jr., and J. R. Weir, Jr., "Development of a Titanium- Modified Hastelloy with Improved Resistance to Radiation Damage," Proceedings of a Symposium on the Effects of Radiation on Structural Metals, San Francisco, Calif., June 23-28, 1968, to be published. R. E. Nightingale, Nuclear Graphite, Academic Press, New York, 1962. J.H.W. Simmons, Radiation Damage in Graphite, Pergamon Press, New York, 1965. R. W. Henson, A. J. Perks, and J.H.W. Simmons, Lattice Parameter and Dimensional Changes in Graphite Irradiated Between 300 and 1350°C, AFRE-R 5489, Atomic Energy Research Establishment, p. 33 (June 1967). C. R. Kennedy, Gas Cooled Reactor Program Semiann. Progr. Rept. Mar. 31, 1964, ORNL-3619, Oak Ridge National Laboratory, pp. 151—-154. 2 J W. C. Riley, High-Temperature Materials and Technology, ed. by I. E. Campbell and E. M. Sherwood, Wiley, New York, 1967, p. 188. w. H. Cbok, Molten-Salt Reactor Program Semiann. Progr. Rept. July 31, 1964, ORNL-3708, Oak Ridge National Laboratory, p. 377. —) L. S. Richardson, D. C. Vreeland, and W. D. Manly, Corrosion by Molten Fluorides, ORNL-1491, Oak Ridge National Laboratory (March 17, 1953). G. M. Adamson, R. S. Crouse, and W. D. Manly, Interim Report on Corrosion by Alkali-Metal Fluorides: Oak Ridge National lLaboratory. Work to May 1, 1953, ORNL-2337, 38 23. G. M. Adamson, R. S. Crouse, and W. D. Manly, Interim Report on ~Corrosion by Zirconium-Base Fluorides, ORNL-2338, Oak Ridge National Laboratory (Jan. 3, 1961). 24. W. B. Cottrell, T. E. Crabtree, A. L. Davis, and W. G. Piper, Dis- assembly and Postoperative Examination of the Aircraft Reactor Experiment, ORNL-1868, Oak Ridge National Laboratory (April 2, 1958). Z25. W. D. Manly, G. M. Adamson, Jr., J. H. Coobs, J. H. DeVan, D. A. Douglas, E. E. Hoffman, and P. Patriarca, Aircraft Reactor Experiment — Metallurgical Aspects, ORNL-2349, Oak Ridge National Laboratory (Dec. 20, 1957), pp. 2-24. 26. W. D. Manly, J. H. Coobs, J. H. DeVan, D. A. Douglas, H. Inouye, P. Patriarca, T. K. Roche, and J. L. Scott, Progr. Nucl. Energy Ser. IV 2, 164-179 (1960). 27. W. D. Manly, J. W. Allen, W. H. Cook, J. H. DeVan, D. A. Douglas, H. Inouye, D. H. Jansen, P. Patriarca, T. K. Roche, G. M. Slaughter, A. Taboada, and G. M. Tolson, Fluid Fuel Reactors, ed. by James A. Lane, H. G. MacPherson, and Frank Maslan, Addison-Wesley, Reading, Pa., 1958, pp. 595-604. 28. Molten-Salt Reactor Program Status Report, ORNL-CF-58-5-3, Oak Ridge National Laboratory (May 1, 1958), pp. 112-113. 29. J. H. DeVan and R. B. Evans III, pp. 557-579 in Conference on Corrosion of Reactor Materials, June 48, 1962, Proceedings Vol. II, International Atomic Energy Agency, Vienna, 1962. 30. J. W. Koger and A. P. Litman, Compatibility of Hastelloy N and Croloy 9M with NaBF;-NaF-KBF, (90—4—6 mole %) Fluorcborate Salt, ORNL-TM-2490, Oak Ridge National Laboratory, in preparation. 39 31. J. W. Koger and A. P. Litman, Molten-Salt Reactor Program Semiann. Progr. Rept. Feb. 29, 1968, ORNL-4254, Oak Ridge National Laboratory, pp. 221-225. . 32. Molten=-Salt Reactor Program Semiann. Progr. Rept. Aug. 31, 1968, ORNL-4344, Oak Ridge National Laboratory, in press. 33. W. R. Grimes, G. M. Watson, J. H. DeVan, and R. B. Evans, p. 559 in Conference on the Use of Radioisotopes in the Physical Sciences and Industry, Sept. 6—17, 1960, Proceedings Vol. III, International Atomic Energy Agency, Vienna, 1962. »— 1-3. =5, 6—15. 16. 17. 18. 19. 20. 21. 22. 23. 2. 25. 26. 27. 28. 29. 30-34. 35. 36. 37. 38. 39. 40. 41. 42. 43, et 45, 46. 47, 48. 49, 50. 51. 52. 53. 5. 55. 56. 57. 58. 59, 60. 61. HEHGORANDOENERUQONPEHGAPNOHNOREENEO0NGOR S D aw 41 INTERNAL DISTRIBUTION Central Research Library ORNL — Y-12 Technical Library Document Reference Section Laboratory Records Department Laboratory Records, ORNL RC ORNL Patent Office Adams Adamson, Jr. Affel Anderson Apple Baes Baker Ball Bamberger Barton Bauman Beall Beatty . Bell Bender E. Bettis S. Bettis S. Billington E. Blanco F 0 E SPHEASEGERAIRER . Blankenship . Blomeke . Bloom Blumberg G. Bohlmann J. Borkowski E. Boyd Braunstein Bredig Briggs Bronstein Brunton . Canonico Cantor Carter Cathers Cavin Chandler Clark Cobb Cochran POPE P OWEmEWH P 62. 63. 6. 65—69. 70. 71. 72. 73. T, 75. 76. 77. 78. 79. 80. g1. 82. 83. 8. 85. g6. g87. g8. 89. 90. 91. 92. 93. . 95. 96—100, 101. 102. 103. 104. 105. 106. 107. 108. 109. 110. 111. 112. 113, 114, 115. PP ORI LR RO NP O N PPN NS PN AP RN LD AP SR e O HENOQTEr NI EONEPE AT P PRI TP NS GE N EOEEE N ROl E SR ORNL-TM-2511 Collins Compere Cook Cook Corbin O ™ Crowley Culler Cuneo Cunningham Dale Davis Davis DeBakker DeVan Ditto Dudley Dworkin Dyslin Eatherly Engel Epler Ferguson Ferris Fraas Friedman Frye, Jr. Furlong Gabbard Gallaher Gehlbach Gibbons Gilpatrick Goodwin Grimes Grindell Gunkel Guymon Hammond Hannaford Harley Harman Harms Harrill Haubenreich Helms 116. 117. 118. 119-121. 122. 123. 124. 125. 126. 127. 128. 129. 130. 131. 132. 133. 134. 135-139. 140. 141. 142. 143, 144, 145-149. 150. 151. 152. 153. 154, 155. 156. 157. 158. 159. 160-164%. 165. leob. 167. 168. 169. 170. 171. 172. 173. 174, 175. 176. 177. 178. 179. 180. 181-185. HEEEEEOO I EII R PHQPIRHQEOCAP TN CGEROR R EENERAP I UOEICY T 42 Herndon Hess Hightower Hill Hoffman Holmes Holz Horton Houtzeel L. Hudson R. Huntley Inouye Jordan Kasten Kedl Kelley Kelly Kennedy Kerlin Kerr Keyes Kiplinger Kirslis Koger Korsmeyer Krakoviak Kress Krewson Lamb Lane Larson Lawrence - Lin Lindauer Litman Llewellyn Long, Jr. Lotts Lundin Lyon Macklin MacPherson MacPherson Mailen Manning Martin Martin Mateer . Mauney McClaln W. MeClung E. McCoy EoREAI=EAQ LSRR RN R IR EPEEAEDEN SO IED O 0D 186. 187. 188. 189. 190. 191. 192. 193. 194. 195. 196. 197. 198. 199. 200. 201. 202. 203. 204, 205, 206. 207. 208. 209, 210. 211. 212. 213. 214, 215, 216. 217. 218. - 219. 220224, 225. 226. 227. 228. 229. 230. 231. 232. 233. 234, 235. 236. 237. 238. 239. 240. 241, CEEPAENOERCEPES TR NERSYR P ES 0O EROPpREIEIHUOORESA ?‘?*?’F’? McElroy McGlothlan McHargue McNeese McWherter Metz Meyer Moore Moulton Mueller Nelms Nichol Nichols Nicholson Oakes Patrlarca Perry Pickel Piper Prince Ragan Redford ichardson Robbins Robertson Robinson Romberger Ross Savage Schaffer . . Schilling CETEPERPOHIEERE MfiQPFQprPergs Seagren Sessions Shaffer Sides Slaughter Smith Smith Smith Smith Smith piewak Steffy Stoddart Stone Strehlow sSundberg Tallackson Taylor erry QP PoEgEERER S E PP EO0 . 242, 243, 244, 245, 246, . 247, 24.8. 249, 250. 251. 252. 253. 254—258. 272. 273. 274. 275. 276. 277, 278. 279. 280. . 281. 282. 283. 284. 285. 286. 287. 288. 289. 290. 291. 292. 293. 29%. 295. 296. 297, 298. 299. 300. 301—315. — 5 a GrPQEoOITEDEYE SREEUSQQUPHEIZQAPIOQDEO D IOENnE * JEIADREOREE R AE BEEEEOO C.') ErQEERSPHED 43 - e P Thoma, ‘ 259. W. J. Werner Thomason 260, K. W. West Toth 261, M. E. Whatley Trauger 262, J. C. White Unger 263. R. P. Wichner Waddell 264, F. W, Wiffen Watson 265. L. V. Wilson Watson 266. J. W. Woods Watts | 267. Gale Young Weaver 268. H. C. Young Webster 269. J. P. Young Weinberg 270. E. L. Youngblood Weir, Jr. 271. F. C. 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