AK R}DGF NATfO‘NAL LABORATORY L M!i i MHB i 3 445k 0548176 O ORNT-40O3T Contract No. W-Th0S-eng-26 MOLTEN-SALT REACTOR PROGRAM SEMTANNUAI PROGRESS REPORT For Period Ending August 31, 1966 R. B. Briggs, Program Director JANUARY 1967 OAK RIDGE NATTIONAL LABORATORY ODak Ridge, Tennessee ocperated by UNION CARBIDE CORPORATION for the U.8. ATOMIC ENERGY COMMISSION K RIDGE NATIONAL LABORA R 3 445L 054417k O SUMMARY Part 1. Molten~Salt Reactor Experiment 1. MSRE Operstions and Analysis The reactor power lewvel wag increased in steps to the maximum at- tainable value of 7.2 Mw. The power limitation wag imposed by the heat- transfer capability of the air-cooled radiator, which was much lower than the design value. The heat-transfer coefficients of the primary fuel-to-coolant-salt heat exchanger were also substantially lower than had been expected. Two periocds, each about two weeks long, of relatively steady operation at the maximum power were achilewved. Agide from the power limitation, the performance of the reactor system was Tavorable. The inherent nuclear stability increased with increaging power as had been predicted. The nuclear poisonlng vy 135%e was only 0.3 to 0.4 of the expected value, apparently because of the presence of circulating helium bubbles in the fuel salt which had not been observed in earlier operations at similar conditions. Zero-power reactivity balances showed a slowly increasing positive anomaly which had reached a maximum of 0.3% 8k/k at shutdown. Radiation heating of the primary-sgystem components was in the expected range, and radiation shielding and containment were adequate. A system ghutdown to remove irradiation specimens from the core, which wag planned after an accumulated exposure of 10,000 Mwhr, was advanced when a catastrophic fallure of one of the main radlator blowers occurred at 7800 Mwhr. In addition to the removal and replacement of the core irradiation specimens and the repair of the main blowers, a number of other maintenance jJjobs were performed in the ensuing shutdown. These included: 1. mechanical and electrical repalr of heat-induced damage to the radiator enclosure, 2. replacement of the particle trap in the main reactor off-gas line, which had developed intermittent high resistance to {low, 3. modification of the treated-cooling-water system to eliminate radiolytic gas, 4. repair of water leaks in the reactor cell, 5 . modification of the component~cooling system to improve reliability, N general improvement of the in-plant electrical system. The MSRE instrumentation and controls system continued to perform well., There was the normally expected reduction in both malfunctions and misoperation of instruments as instrument and operating personnel gained experience and developed routines. While there were many design changes, most of these were improvements and additiong to the gystem rather than correctlive measures to the instruments and contrcls. A iii iv disappointingly large number of faulty commercial relays and electronic switches were disclosed. These faults were in the areas of both relay deslgn and fabrication, and corrective steps have been taken. 2. Component Development The operation of freeze valve FV-103 was improved by the deletion of the hysteresis feature of one of the temperature-control modules, which had caused thermal cycling of the wvalve before the ftemperature reached equilibriunm. The three conlrol rods have operated without difficulty. The indiecated changes in rod length were random and were 0,05 in. for rod 1, 0.16 in. for rod 2, and 0.12 in. for rod 3. Several small difficulties were encountered in the operation of the control-rod drive units. In each case the difficulty was diagnosed and adequate temporary changes were made to permit contiunued operation of the reactor. The difficulties involved the failure of a synchro transmitter and a reference potentiometer, which have been replaced during the current shutdown. The excellent condition of the gears in the drive units indicated that there was no high-temperature damage €O the lubricant. The radiator-door-operalting mechanism has performed satisfactorily since the last modifications. Excessive alr leakage around the seals ori the doors resulted from damsge sustained during the thermal cycling during normal operation. Laboratory tests were conducted on several arrangements of the metal seal surfaces, resulting in the choice of a new hard-gceal scheme which was installed on the door. In addition, alterations were made to the door structure to reduce bowing, and the installation of a second soff,; resilient seal was proposed to back up the existing seal. The sampler-enricher has been used to isolate a total of 119 10-g samples and 20 50-g samples and to make 87 enrichments to the fuel sys- tem. The only major maintenance required has been replacement of the manipulator boots, replacement of the drive-unit capsule latch, repair of an open electrical circuit, and recovery of a capsule which had fallen into the operational valve area. These repairs were performed without the spread of airborne contamination or the exposure of the personnel to significant radiation levels. An increase of the buffer leakage in the operational valve was determined to be caused by particles which fell onto the upper seal surface, and it was found that simple cleaning of this surface was effectlive in reducing the leakage. Problems resulting from an increase in the contamination level within the mech- anism were solved by a partial cleanup of one aresa, the establishment of a contamination control area at the sample withdrawal arca, and the modification of the transport container to reduce contamination of the upper part and to permit inexpensive disposal of the lower part. Changes were made in the sampler~enricher control cireuit to reduce the chance of rupturing the manipulator boots. In addition, a fuse and voliage suppressor were installed to protect the electrical leads {from excessive voltages and currents. A total of 45 samples, including two 50-g samples, have been taken using the coolant sampler. One electrical receptacle was replaced, and the lezkage of the removal wvalve wag reduced by cleaning. The design of the fuel processing sampler was essentially completed, and installation i1s proceeding as craflt is avallable. The electrical and instrument work is about 50% complebe, and the installation of other equipment is over 95% complete. The original off-gas Tilter in line 522 was replaced with one which 1s designed to trap organic materials in addition to particulate matter. Activated charcoal was chosen as the filtering medium, and preliminary tests indicated that 1t had good efficiency for the removal of Cg and heavier molecules. A prefilter was installed to remove the radicactive particulate matter as well as the organic mists which might exist. Since one of the purposes of the prefilter was to reduce the heat load on the charcoal trap, heat dissipation was also a consideration in the design. The entire filter—charcoal trap is cooled by immersion in watey. Bince little was known of the character of the crganic ma- terial at the time of the design of the filter system, one of the pur- poses was to provide a method of diagnosing the problem. Therefore, the particle trap will be removed and examined to gain information needed in the design of a more permanent system. Diffusion of activity into the fuel pump off-gas line resulted in two short periocds of high activity at the off~-gas stack. A small char- coal trap was installed to provide holdup times of approximately 2-1/2 days for krypton and 30 days for xenon. Several remote maintenance jobs were performed during the periocd, in which the accumulated reactor power increased fTrom 35 to 7822 Mwhr. Several observations were made as a result of the work. Control of air- borne contamination is not difficult, and the maintenance techniques and systems prepared for the MSRE have worked well. The flexibility of the maintenance approach was demonstrated by carrying oul unanticipated tasks such as the installation of a thermocouple on a piece of pipe in the reactor cell and the thawing out and clearing of a plug in one of the service gas lines. The major tasks completed during the report in- clude: (1) opening a section of the off-gas line, inspecting the inside, and returning the line to operating condition; (2) removal and replace- ment of both of the reactor-cell space coolers; (3) installation of = new thermocouple on a horizontal seciion of the off-gas line; (4) repair of the sampler-enricher electrical receptacle; (5) installation of tem- porary piping in the main off-gas line to measure pressure distributions; (6) removal of the graphite—flastelloy N surveillance samples; (7) re- moval, repair, and replacement of two control-rod drive units; (&) re- moval of a salt plug from & gas~pressure reference line by applying pressure while heating the line. Vi 3. Pump Development The MSRE prototype fuel pump was operated for 2631 hr at 1200°F to obtain data on the concentration of undissolved helium in the circulating salt and on the hydrocarbon concentration in the pump-tank and catch- basin purge gases. The spare rotary elements for the MSRE fuel and coolant pump and the MK-2 Tuel pump were modified with a seal weld between the bearing housing and shield plug to prevent oil from leaking out of the leakage catch basin and down the outside of the shield plug to the pump-~tank gas space. A lower shaft seal fallure was experienced during preheat of the spare rotary element for the MSRE fuel pump in preparation for hot shakedown. The spare rotary element for the MSRE coolant pump was given cold and hot shakedown tests. The lubrication pump endurance test was continued, and fabrication of the MK-2 fuel pump tank was begun. Operation of the PK-P molten-salt pump was halted by failure of the drive motor. (In the summary section of a previous Progress Report, the number of operating hours was reported in error as 22,622. Total for four tests is 23,426 hr.) The gimbals support for the salt bearing on the molten-salt bearing pump was modified, and a new bearing sleeve and journal were fabricated. 4. Instrument Development Performance of the temperature scanning system continues to be satisfactory, although some problems were experienced with oscilloscopes and mercury switches and some system instability was noted. Because spare parts for the mercury switches used in the scanner can no longer be obtained from the manufacturer, an effort is being made to find a replacement for the mercury switch. Further testing of the coolant-salt system flow transmitter which failed in service at the MSRE confirmed that refilling the transmitter with silicone oil had significantly reduced its temperature sensitivity. The new NaK-filled differential-pressure transmitter ordered for use as an MSRE spare was found to be excessively sensitive to pressure and temperature variations during acceptance testing. Performance of all molten-salt level detectors installed at the MSRE, on the MSRP level Test Facility, and on the MSRE Prototype Pump Test Loop continues to be satisfactory. To correct excessive frequency drift present in the excitation oscillator supplying the ultrasonic lewvel probe, a number of minor changes in components and circuitry were made in electronic equipment associated with the probe. Modification and/or repalir of two defective helium control valves was completed. The feasgibility of using sliding disk valves for control of very small dry-helium flows is being investigated. 5. Reactor Analysis Rod drop experiments, performed during MSRE run No. 3, were analyzed and compared with rod worths determined from other independent measure- ments. Theoretical time~integrated flux trajectories following rod scrams were calculated, based on negative reactivity insertions obtained by integrating differential worth measurements. These trajectories were found to compare closely with experimental reccrds of the accumulated count following the scram. We have concluded that an approximate 5% band of self-consistency can be assigned to the control rod reactivity worths inferred Irom these two independent calibration technidues. Part 2. Materials Studies 6. MSRP Materials The grade CGB graphite and Hastelloy N specimens were removed from the core of the MSRE after 7800 Mwhr of operation. Thelr macroscopic appearances were essentially wunchanged by this exposure. Some of the specimens were damaged physically as a result of differences in thermsl expansion of parts of the assembly. A new core gpecimen array was as- sembled with modifications to correct these difficulties. A metallurgical investigation was conducted to determine the effect of aluminum-zinc alloy contamination on the Hastelloy N tubing of the MSRE salt-to-ailr radiator. Contamination occurred from a blower fallure during which shrapnel was blown across the hot radiator tubes. ILabora- tory tests showed that, in general, an aluminum oxide coating contained the aluminum, even in the molten state, and interaction did not occur. When the oxide skin wag broken from mechanical abrasion, shock, or other reasons, wetting occurred. Moderate interaction to a depth of about 0.010 in. occurred in a wetted sample held at 1200°F for 5 hr. The tubes in the radiator were inspected, and those which were contaminated were carefully cleaned. As a result of the investigation and cleanup pro- cedure, the radlator system was Judged to be satisfactory for further operation. ; Examinations of new grades of both anisotropic and isotropic graphite indicate that these do not yet meet the reguirements of molten-salt breeder reactors. Results of a variety of graphite creep experiments performed over a wide temperature range support a Cottrell model for irradiation creep. Use of this model will permit easier exbrapolation of data. Implied in this concept is the conclusion that as long as the stresgs acting on the graphite does not exceed the fracture stress, the graphite will continue to absorb the creep def'ormation without loss of mechanical integrity. The experimental brazing alloy 60 Pd—35 Ni—5 Cr (wt %) was evaluated for joining graphite to metals. Although it exhibits relatively poor wettability on high-density graphite, 1its marginal behavior is enhanced by preplacing it as foll in the joint. Graphite-to-molybdenum joints brazed with this alloy preplaced in the joint were thermally cycled between 200 and 700°C, and metallographic investigation showed that no deterioration had occurred. Two graphite-to~-molybdenum~-to-Hasgtelloy-N viii transition joints were brazed using a tapered joint design. Visual examination revealed no cracks, and evaluation i1s continulng. A new brazing alloy, 35 Ni—60 Pd—5 Cr (wt %), for joining graphite to molybdenum had less than 2 mils attack after exposure to LiF-~-Belp- ZrF,~ThF,~-UF, at 1300°F for 5000 hr in a Hastelloy N container. Since zirconium and titanium have been found to improve the resis- tance of Hastelloy N to effects of neutron irradlation, the 1nfluence of these elements on the weldability is being evaluated. Titanlium ap- pears to have no deleterious effects. Zirconium in concentrations as low as 0.06 wt % causes hot cracking. However, reasonably good welds have been made by the use of filler wire that contains dissimilar metal. The level of zirconium that can be tolerated has yet to be determined, Our studies of the behavior of the Hastelloy N under neutron irra- diation have been concerned with evaluating the heats of material used in the MSRE and in evaluating several modified heats of Hastelloy N for use in an advanced system. In-reactor stress-rupbure tests on several heats of material suggest that there may be a stress below which es- sentially no neutron damage occurs. Tests on specimens exposed to var- ious ratios of fast flux to thermal flux indicate that the dsmage cor- relates with the thermal flux. We believe that the damage 1s due to helium produced hy the (n,&) reaction with 98, We have produced several heats of material with very low boron, but have had to change from air to vacuum melting practice. If the low-boron material is irradiated cold, we find that the properties are superior to those of the higher boron material; if irradiated hot, the properties are as bad or worse. Thus the postirradiation properties are not uniquely dependent upon the boron content, and other factors such as the distribution of boron and the presence of other impurities must be very important. The addition of titanium to Hastelloy N has been very effective in improving the prop- erties. Limited creep-rupture tests have been run on Hasteloy N to deter- mine its suitability for use as a distillatlon vessel for molten salts at 282°C. This work has resulted in a determination of the strength propertice to times of 1000 hr. The formation of a second phase was observed that may influence the ductility at lower temperatures. The oxidation characteristics under cyclic temperatures remain to be deter- mined. Thermal convection loops containing Tused salt of MSRE composition and fabricated from Hastelloy N and type 446 stainless steel have con- tinued cperation for 4.5 and 3.2 years, respectively, with no sign of difficulty. A slight decrease in cold-lcg temperature has been noted in an Nb—1% Zr loop after 0.6 year. A Hastelloy N loop has circulated a secondary coolant salt for 3C00 hr, whereas a Croloy 9M loop with the same salt plugged in 1440 hr. 7 Chemistrz The fuel and coolant salt have not changed perceptibly in composition gince they were first circulated in the reactor some 16 months ago. The ix concentration of corrosion products has not increased appreciably. The average oxide concentration in the fuel was 54 ppm, which is reassuringly low. The viscosgity and density of molten Bels were measured; the vis- cosity was about 10% greater than previously reported, and the density of the ligquld is not very different from that of the solid. Vapor eguilibria that are involved in the reprocessing by distil- lation have been measured. Decontamination factors of the order of 1000 Tor rare earths were evidenced. Thermophysical properties have been estimated for the sodium potas- aium fluoroborate mixture that dis a proposed coolant for the M3BR. The vapor pressure, due to evolution of BF;, reaches 229 mm at the highest operating temperature. Interim estimates for density, specific heat, and viscosity of the proposed coolant were made available. Possible reprocessing methods were studied 1ln greater detail. Fun- damental studies related to the thermodynamics of the reduction of fission product rare earths into a bismuth alloy were carried out. The exceed-~ ingly low activity coefficients of rare earths in the bismuth explained the feasibility of the process. Further attention was pald to the re- moval of rare earths by precipitation in a solid solution with UFs. The removal of protactinium from blanket melts was studied in sev- eral ways. These included an oxide precipitation with ZrO,, a pump loop to transfer protactinium in a bismuth-thorium alloy, and attemptes at electrolytic reduction from blanket melts. Moderate success was achleved in these experiments, but more work is required to arrive at finished and fully controlled procedures. Analyses obtained from sampling assemblies that had been exposed in the pump bowl of the MSRE showed thal noble-metal fission products were being partially released to the gas space, presumably asg volatile fluorides. At the same time, plating of noble metals from the liquid was encountered. These puzzling phenomena were reflechbed in results on surveillance specimeng of graphite and metal which were removed from the MSRE. Some 10 to 20% of the yield of noble-metal fission products was found to have entered the gas space in the graphite. Analyses of xenon isotope ratios in concentrated samples of off-gas from the MSRE showed that the burnup of 135%e wags about 8%; the remainder escaped to the cover gas or decayed. This is in accord with the low xenon polsoning indicated by reactivity behavior. Preliminary estimates of xenon poisoning and cesium carbide for- mation in the MSER indicate that cesium deposition will probably not be a serious problem, but that stripping for iodine removal will probably be required Lo keep poiscning within bounds. Oxide concentrations of 50 to 70 ppm were determined in fuel samples taken from the reactor duriag operations at all power levels without apparent interference from the activities of the samples. Techniques for the regeneration of elec- trolytic moisture cells were developed to provide dependable replacementg for the hot-cell oxide apparatus and components for future in-line ap- plications. Measurements directed toward the development of in-line speclro- vhotometric methods disclosed additional wavelengths of potential analyt- ical value in the ultraviolet absorption spectrum of U(III) and confirmed the absence of interference from corrosion products. Investigation of unusual valence states of rare-earth fission products indicates possible interference from Sm(IT) but none from Eu(Il). An intease absorption peak suitable fTor monitoring traces of uranium in coclant salt has been found in the ultraviolet spectrum of U(IV). A modified optical system has been ordered which will improve the spectrophotometiric measurcments of molten Tluoride salts. By voltammetric and chronopotentiometric measurements, the U(IV) reduction wave was found to correspond to a one-electron reversible re- action. Diffusion coeflicients and the activation energy were measurcd. Repeated scans of this wave in gquiescent MSREE wmelts were reproducible to about 2% over extended periods and better than 1% during short-term measurements, A new voltammeter is being bulilt to improve the reproduc- ibility and make possible measurement of flowing salt streams. Design criteria are being considered for an in-line test facility for evaluating three types ol continuous analytical methods. Hydrocarbons were measured in helium from simulated pump leak ex- periments, and an apparatus was developed for the continuous measurement of hydrocarbouns in MSRE off~-gas. Efforts were continued on the development and evaluation of equip- ment and procedures for analyzing radioactive MSRE salt samples. The remote apparatus for oxide determinations was installed in cell 3 of the High~-Radiation-Tevel Analytical Laboratory (Building 2026). In addition to the analyses performed on the salt samples, radio- chemical leach solutions were prepared on silver and Hastelloy N wires coiled onto the stalnless steel cable between the latceh and ladle. The quality-control program was continued during the past period to establish more realistic limits of error for the methods. Part 3. Molten-Salt Breeder Reactor Studies 8. Molten-Zalt Breeder Reactor Design Studies Further design changes were incorporated into the reference molten- salt breeder reactor concept. The design of the primary heal exchangers wag altered to eliminate the need for expansion bellows, Also, the flow of fluid in the primary reactor circuits was reversed to lower the oper- ating pressure in the reactor vescel. The effect of lowering the feedwater temperature from 700 to 580°F was evaluated. Tt was found that this change increased the plant thermal efficiency from 44.9 to 45.4% and reduced plant construction costs by $465,000 if there were no accompanying adverse effects. These savings are canceled if the coolant used with the lower feedwater temperature costs $2.4 million more than the coolant used with 700°F feedwater. Molten-salt reactors appear well suited for modular-type plant con- struction. Such construction causes no significant penalty to either X1 the power-production cost or the nuclear performance, and it may permit MSBER's to have very high plant-availability factors. Use of direct-contact cooling of molten salts with lead significantly improves the potential performance of molten-salt reactors and indicates the versatility of molten salts as reactor fuelsg. However, in order to attain the technology status required for such concepts, a development program 1s necessary. The molten~galt reactor concept that redquires the least amount of development effort is the MSCR, but 1t is not 2 breeder system. The equilibrium breeding ratio and the power-production cost of the MSCR plant were estimated to be about 0.96 and 2.9 mills/kwhr (electrical), regpectively, in an investor-owned plant with a load factor of 0.8. Although this represents excellent performance as an advanced converter, the development of MSBR(Pz) or MSBR plants appears preferable because of the lower power-production costs and superior nuclear and fuel-conser- vation characteristices associated with the breeder reactors. 9. Molten-8alt Reactor Processing Studies The processing plant for an MSBR would use side streams withdrawn from the fuel- and fertile-salt recirculating systems at rates that vield a fuel-salt cycle time of approximately 40 days snd fertile-salt cycle time of approximately 20 days. Among the significant steps in the presently envisioned process are recovery of the uranium by continuous fluorination and recovery of the carrier fuel salt by semicontlinuous vacuum distillation. Alternative schemes are also belng conszidered. Semicontinuous Distillation. Values of the relative volatilities of NdF;, Lal;, and CeFs; in LiF are of the order of 0.0007. These are from new measurements made using a recirculating equilibrium still. Barlier measurements made by a cold-finger technique were zbout a factor of 50 too high. The complexity of still design and operation is consid- erably eased by these lower values. Retention of over 209 of the rare-~ earth neutron poisons in less than 0.5% of the processed salt can easily be achieved. Continuous Fluorination of a Molten Salt. The uranium in the fuel stream of an MSER must be removed by continuous fluorination pricr to the distillation step. The significant problems are corrosion of the flucrinator and the possible loss of uranium. Shtudies are in progress on continuous fluorinators constructed as towers with countercurrent flow of fluorine to salt. Recoveries exceeding 99% have been consistently attained with towers only 48 in. high. Higher recoveries with longer towers are anticipated. Corrosion protection may be effected by the use of a layer of frozen salt on the wall of the fluorinator. Feasibility of this technique is based on successful experiments with batch systems and simple heat transfer calculations. The heat generation oif the fuel salt should be adequate to maintain an easily controlled layer of frozen salt on the cooled metal wall. Alternative Chemical Processing Metheds for an MEBR, Reductive co-~ precipitation and liquid-metal extraction are being studied as possible x1i methods for decontamination of MSBR carrier salt (LisBeF,) after uranium removal by the fluoride volatilization process. Adequate removal of La and Gd. is achieved by treatment with near-theoretical quantities of beryl- lium metal to form beryllides of the type InBeq,. Either excess Be, up to 243 times the theoretical amount, or a stronger reductant, Li, is nec- essary Lo remove zirconium at trace level. The zirconium is removed as free metal from 5 mole % ZrF, solutions in LipBeF,, but from dilute solutions a beryllide, ZrBes, has also been identified. Lithium~-bismuth liquid-metval extraction experiments were also con- tinued. Significant removals were observed for La, Sm, G4, Sr, and Eu, the latter principally by extraction into the metal phase, the others by deposition as interface solids, as previously reported for other metal extraction tests. CONTENTS Sm....l..ll..."..lll.li.!..tl....l‘.l" ------ > a & IMRODUCTION.......Q......I.-.O...-’I.l.........'.-...-n'.fllifll'.. l Part 1. MOLTEN-SALT REACTOR EXPERIMENT 1. MSRE OPERATTIONS AND ANATLY SIS.eseessessassscscsecsassasnsanancs ' 1.1 Chronological Account of Operations and Maintenance..... 7 1.2 Reachivity BalanCe.eiessnesesnsssssrsssancassassnsscssnns 10 Reactivity Balances 8 POWer.ceeesseecacscerarssannons 11 Reactivity Balances at LOWw PoWwer..seeacscceccscnnense 12 1.3 Nenon POlcONingeesessocsssoccesnacossrrssssernsasnssnssns 13 Predicted Steady-State 12°Xe PoisOning.cevsseessssees 14 Analysis of Transient *2°Xe Poizoning..e.esseeeess--o 16 4 Cireulating Bubbles.sseveseseeerons ceeearrreaanaannn ceus 22 5 Salt TranspoTt.eeseesscecocrsaecane cessesccacscsarnuaeae 2% Gradual Transfer Lo Overflow Tank....ceeeeesvescscess 24 OVEYTIi]l ) s ssoenasnssssasseansssnnarsassssossnsenasosssan 24 1.6 Power MeasurelenlS.ieeesenosassasssscaasasassonnn venesane 25 Heat BalolCO.seeesssscscassascssssossonsasssssssaansanas 25 Muclear INstrumentSseesessssecosssscssessossannesas . 26 Radiator ALlr FlOW.sesesseraanase frsesesntseasaessennes 26 1.7 Radiation Heabting..eeeeessceaacsssasserssossssssassssenns 27 Fuel-Pump Tank.eseeeseessae breees chessaesasssrssenb e 27 Reactor Vessel.eerneessrasssnaneas ceneaen eraaeas coees 28 Thermal Shield.ecseeacsssnssocsnssssvscscsaasvocassnnss 29 1.8 Reachor DynamicS.ceeessesesssscsssssssssrsssasenessassss 29 1.9 TEquipment Performance....ceeseasossssversscessonsocconss 35 Heat Trancleri.ieecscossasseans cenaa teeceesersenacnse 35 Main BlOWerSeesecescassorsocnsonsnas tessssasnesnerses 39 Radiator ENcloSUl€.casssevssesssasssasssssncssonssenssens 4. Fuel Off-Gas Syslem..eescaeeassceorsesssseasnsoasnacs 43 Trested Cooling-Water System....eiieiecesercacsrsonns 47 Component~Cooling Systell.ceesssoscoassassnsascascsns . 48 Salt~Pump O1l SySTEmMS: ccveasessansasssesanccasssasnss 50 Electrical System...ceeecevoeas pacesrarsacarvasnaseans 51 Control Rods and Drives.se.sesesrasosnnsansss e aamesee 53 SaMP LB S e e aaessssssosssescasnassasssseassesssassersssss 53 Contalnment .. oeescessseressssassassasnsnnssssscnnnnas 54 Shielding and RadiafioNeesascesescsrssonsaas cascasaas 56 1.10 Instrumentation and ControlS..ssesessscscssssassesasonss 57 Operating Experience — Process and Nuclear Instbruments.cesesenoscaaeasvas eessesarnsnoasrennnee 58 Daba SyShell.eeosenvssasnasesnscsssarsasoscecssssssnss 59 Control-System Designe cesosceassssasososssaasasassnas ol 2& COD’[PONEN‘I]:‘ DEVELOPIVEIq’mI‘l...l..fll..i.0‘t'fl.l..l...fl.l'i.flll'ln' 6-') xiii Xiv 2.1 Ireeze ValveS.ieaeeeieeersosssetseesstsscscnecssascnceacsscse 65 2.2 Control ROUSeecesesseasessssocsssscaasosacscsosassosossseasse 65 2e3 Control-Rod Drive UnitSeeeeesececsscsescssoasosssnossosesss 66 2od RAdiatior DOOT Sessesssssssrsssssssarssassasscssscvcssssaess o7 265 oD e -l GO s aseasseesasossessrsssssssnscsscsssnanssa 70 Replacement of Manipulator BOOLS.ssssssscsssesacsanna 70 Replacement of the Capsule LabCh eeeeseesesresccerons 71 Repair of an Open Electrical Circuit..sceececscsascas 7L Recovery Of a CapslUlCasscesessossracsssscassecesansascea 72 Cperational Valve 1eakige.sececcscasscssasnosnaconcea 72 Contamination of Removal Valve Se8lSeiecescasscrssonsans 72 Miscellaneous ProblelS.ceecescessessserserssacrersannss 73 Changes to the Control Circuit..eceeeesecsacsoesasses 73 Coolant SampPleTesesessaessscsssesrssssssvsossscossrsonsssns 73 Fuel Processing System Sampler.sssecscscescacessnsansans 7 Off~Cas Tllter Ascemblyeseecesesoastassrsesorscaccssnessse "4 Filter MediUMeceesesssssasossrascassasseanacsnssessnaa T4 Heat Diceipalion,. caeessseescsssssaossososssntrsoncroases M Particle TrDesscesessarsasscsaccasssssscoossesssscesess 75 Charc08l TraPeersscccsensssscecscscssesssanssessacssessss 76 2¢9 524 Charcoal BeQeeessssscsscsscassssssssnesnssnsssnsssossos 7 2.10 Remote MalntenanCE.ceeessesccsssessesessoaenssnasosansena 77 oMM o -3 O Pm@ .DEVEIJOP}JENT..."....'...‘..-.........-....I............l. 81 3.1 MORE PUlDSesessasesasoscsascssansscsssssosscesnastoscnsossanns g1 Molten~-Salt Pump Operation in the Prototype Pump Test Facllifyeeesstssesaessassesscanssssnscacsncnss 81 Pump Rotary Element Modificabtions.esesseecesscceasoas 82 Iubrication SysteMececcesssssesssssessaseesssssnnesss 82 MK=~2 Fuel PUllDessessssvssscssasnsssconssensecscancecses 82 3.2 Other Molten-8alt PuMpPSeesacscececssescavoasososansssons 82 PK-P Fuel-Pump High~Temperabture Endurance Teot.eeeee. 82 Pump Containing a Molten-Salt-Lubricated Journal Bear N escsaessessessssnessesnsassssncscsasnssssasans 82 INSTR[M]NTDEVELOPI%]NT...II.‘..C.....I..I..lDl...l.ll..l..'..l 84‘ 4. 1 Ten‘.l)e 1‘.atux.e Sc al‘l-rler * o 5 o0 PSR PSR RS E T S e ST P e R e G S e e R 8ZE' 4e2 High-Temperature, NaK-Filled, Differential-Pressure TransmittEr. 8 5 0 8 9 0 PR e SO TSrE S PO S RSP E SRS SN RS S e S 85 lee 3 Molten~3alt Tevel DeteCtOrSeeessosccssassessosscccssssssa 85 doet Helium Control Valve Trim Replacellenl.eesesececasccosess 86 mACTORANA:LYSIS.-..'...‘......I....‘....‘....l.........'.-..l 88 5.1 Analysis of Rod Drop ExperimentS.cecececsrsenssoacsacsnss 88 Description of EXperimentS.eeeeeececscesscscoseasossas 88 Analysis ProCcelUreS.esessecssassarssssscsarscssassanas g9 RESULllSeesevseasessansssncssecsassocasacsasassarsnnsana 91 6. I&SI%P D&aterials.....“...I-I..‘D.ll 6.1 .2 o Oy W SOy Oy - ~3 O\ 6.8 XV Part 2. MATERIALS STUDIRS 4 & 2 8 8 9% ¢ 90 & 00 S P DI R ST S A B MERE Materials Surveillance Testing.sceesvessasssessacos Evaluation of Possible MSRE Radiator Tubing Contamination with Aluminum,.. 2 9 o % 00 2 # & 2 2 @ % 3 2 2 & ¥y 00 > TR Evaluation Df Graphitei..'I.I...C'.I.ll...l‘lllD‘.ll..l. Internal Stress Problem in Graphite Moderator Blocks...llI‘...ll..lll...'hfll Brazing of Graphite.....cssevces 480 0 A8 ST S sERA BB F R RaE Corrosion of Graphite~to-Metal Brared Jointo.i.eeeeeoveas Welding Development of Hactelloy Nesecessseassascsaonasas ffect of Irradistion on the Mechanical Propertics of Hastelloy Neeseseseescnsscorscsocnsco Characterivation of Hastelloy N for Service At F82% Ciernnenonennnsarucnansasossnnsnsossnnassasss Thermal Convectbion Loops...........,.................... r?. CI{EMISTI{I"....’..."...I....."....‘0............-l’i‘....’fi..‘ 7el 7o 7.3 77- ZI’ 765 Chemistry of the MORE. eeevecoacsassssssnsnsnconssnsnass Behavior of Fuel and Coolant Salt.sssesesesvacsssnces Physical Chemistry of Fluoride Mellt.eeeeeesersseancanes Viscosity and Density of Molten Beryllium F:Lu.oride-p..---o..--oaoounn ------ Transpiration Studies in Support of the Vacuum DiStillationPI‘OCGSS.'..IO......B...III.I..'O..O..’ Estimated Thermophysical Properties of MSBR Coolan-t Salt‘.......l...-‘. & & 9P oS s S PR e T eSS Separation in Molten FluorideS.sasseccsssessescsssssccee Extraction of Rare Tarths {rom Molten Fluorides 1nto Molten MebtalSeeeesossesonssassssnasass-snnesnss Removal of Rare Earths from Molten Fluorides by Simultaneous Precipitation Wwith UFesesscsnsssssaas Removal of Protactinium from Molten Fluorides by Oxide Precipitationesscecsass & ® ¢ g » o & 5 & 0" a0 a8 9 & Extraction of Protactinium from Molten Fluoridesgs into Molten Mebtialo.ieearsesesoessssnnoacsnmesssnacass Protactinium Studies in the High-Alpha Molten- Salt Laboratoryesescaseessssscrsacsorssssscasnesssoa Radiation Chemistry.iesereaeesonsocsncssesasrsssassarsnans Xenon Diffusion and Possible Formation of Cesium Carbide in an MSBReesessesaoreesa Fission Product Behavior in the MORE..eeesreesossovas Development and Evaluation of Analytical Methods Tor Molten-5alt ReaC b0 e eerassssocescsesssaracsnseres Determination of Oxide in Radicsactive MSEE * 0 4 8 & F 0 88 S A S e 0 ad SampleSO.C"l....l..'.l’l."..ll...ll.l'...'.@fl!.-. Spectrophotometric Studies of Molten-3alt Reactor mels....ll.....,.l...l.-l. ...... ® & * g % b &9 S5 @0 e s P 9'7 o7 103 108 110 112 115 117 117 125 130 134 134 134 139 139 140 143 142 142 145 147 148 156 158 158 165 191 191 193 740 xvi Voltammetric and Chronopotentiometric Studies of Uranium in Molten LiF-BeFy-ZrF,. In-Line Test Faclilityeeeareoeecans Analysis of Helium Blanket Gas.... 2 5 & 8 % &8 P & s 2 Development and Evaluation of Equipment and Procedures for Analyzing Radiocactive MSRE Salt Samples..ieeseasss Sample AN8lySeS.isesssssossvnsansonss Q,ualitwaO'fltl"Ol PI‘OgI‘Elm. « 8 0P 08 s e Part 3. MOLTEN~SALT BREEDER BEACTOR STUDIES MOLTEN-SALT BREEDER REACTOR DESIGN STUDIES..s.e.. 8.1 8.2 8 8. ™~ W Design Changes in MSBR Plant.....eeee Modular-Type Plant..eeeseessessssonese ¢ 8 2 & ¢ o9 5 0 0 b e oteam Cycle with Alternative Feedwater Temperalure...... Additional Design ConceplSeesscscesas MOLTEN~-SATT REACTOR PROCESSING STUDIES..... 9.1 9.2 9.3 Semicontinuous Distillations.eeeeceess Continuous Fluorination of a Molten Salt.. » = » 9 8 Alternative Chemical Processing Methods for an MSBR..... Reduction Precipitation.c.ceeeeess Li_Bj_ A.lloy EXJGl"aCtiOIl. & % 48848 s 195 196 19 199 200 200 207 207 212 217 223 227 228 232 233 235 237 INTRODUCTION The Molten-Salt Reactor Program is concerned with research and de- velopment for nuclear reactors that use moblle fuels, which are solu- ticns of fissile and fertile materials in suiltable carrier salts. The program is an outgrowth of the ANP efforts to make a molten-salt reactor power plant for aircraft and is extending the technology originated there to the development of reactors for producing low-cost power for civilian uses. The major goal of the program 1s to develop & thermal breeder re- actor. Iuel for thils type of reactor would be 233UF4 or 235UF4 dissolved in a salt of composition near 2LiF-BeF,. The blanket would be Thl, dis- solved in a carrier of similar composition. The technology veing devel- oped for the breeder is applicable to, and could be explolted sooner in, advanced converter reactors or in burners of fissionable uranium and plutonium that also use fluoride fuels. Solutions of uranium, plutonium, and thorium salts in chloride and fluoride carrier salts offer attractive possibilities for mobile fuels for intermediate and fast breeder reactors. The fast reactors are of interest too, but are not a significant part of the program. Our major effort is being applied to the cperation and testing of a Molten-Salt Reactor Experiment. The purpose of this experiment 13 1o test the type of fuels and materials that would be used in the thermal breeder and the converter reactors and to obtain several years of ex- perience with the operation and maintenance of a small molten-~salt power reactor. A successful experiment will demonstrate on a small scale the attractive features and the technical feasibility of these gsystems for large civilian power reactors. The MSRE operates at 1200°F and at atmospheric pregsure and was intended to produce 10 Mw of heat. Initially, the fuel contains 0.9 mole % of UF,, 5 mole % ZrF,, 29.1 mole % Bel'y, and 65 mole % LiF, and the uranium is about 30% “2°U. The melting point is 340°F. In later operation, we expect to use highly enriched uranium in the lower concentration typical of the fuel for the core of a breeder. In each case, the composition of the solvent can be adjusted to retain about the same liquidug temperature. The fuel circulates through a reactor vessel and an external pump and heat-exchange system. All this equipment is constructed of Hastelloy N,l a new nickel-molybdenum-chromium alloy with exceptional resistance to corrosion by molten fluorides and with high strength at high tempera- ture. The reactor core contains an assembly of graphite moderalor bars that are in direct contact with the Tuel. The graphite 1s a new material? of high density and small pore size. The fuel salt does not wet the graphite and therefore should not enter the pores, even at pressures well above the operating pressure,. 1a1s0 sold commercially as Inco No. 806. 2Grade CGB, produced by Carbon Products Division of Union Carbide Corp. Heat produced in the reactor is transferred to a coolant salt in the heat exchanger, and the coolant salt is pumped through a radiator to dissipate the heat to the atmosphere. A small facllity is installed in the MSRE building for occasionally processing the fuel by treatment with gaseous HF and Fs. Design of the MSRE was begun early in the summer of 1960. Orders for special materials were placed in the spring of 196]. Major modifi- cations to Building 7503 at ORNL, in which the reactor is installed, were started in the fall of 1261 and were completed by January 1963. Fabrication of the reactor eguipment was begun early in 1962. Some difficulties were experienced in obtaining materials and in making and installing the equipment, but the essential installalions were completed so that prenuclear testing could begin in August of 1964. The prenuclear testing was completed with only minor difficulties in March of 1965. Some modiflcations were made before beginning the critical experiments in May, and the reactor was first critical on June 1, 1965. The zero- power experiments were completed early in Jduly. Additional modifications, maintenance, and sealing and testing of the containment were required be- fore the reactor began to operate at appreciable power. This work was completed in December. Operation at a power of 1 Mw was begun in January 1966. At that power level, trouble was experienced with plugging of small ports in the control valves in the off-gas system by heavy liquid and varnish- like organic materials. Those materials are belileved 1o be produced from a very small amount of oil that leaks through a gasketed seal and into the fuel salt in the pump tank of the Tuel circulating pump. The oil vaporizes and accompanies the gasecus fission products and helium cover gas purge into the off-gas system. There the intense beta radiation from the krypton and xenon polymerizes some of the hydrocarbons, and the products plug small openings. This difficulty was largely overcome by installing an absolute filter in the off-gas line ahead of the control valves. The full power capability of the reactor — about 7.5 Mw under design conditions — was reached in May. The power was limited by the perfor- mance of the salt-to-alr radiator heat-dump system. The plant was oper- ated until the middle of July to the equivalent of =zbout one month at full power when troubles with the blowers in the heat-dump system re- quired that the operation be Iinterrupted for maintenance. In most respects the reactor has performed very well: +the fuel has been completely stable; the fuel and coolant salts have not corroded the Hastelloy N container material in several thousand hours at 1200°F; and there has been no detectaple reaction between the fuel salt and the grapnite in the core of the reactor. Mechanical difficulties with equip- mentl have been largely confined to peripheral systems and auxiliaries. Because the MSRE 1s of a new and advanced type, substantial research and development effort is provided in support of the operation. Included are engineering development and testing of reactor components and sys- tems, metallurgical development of materials, and studies of the chemistry of the salts and theilr compatibility with graphite and metals both in- pile and out-of-pile. Work is also being done on methods for purifying the fuel salts and in preparing purified mixtures for the reactor and for the research and development studies. Some studles are being made of the large power breeder reactors for which this technology is belng developed. This report is one of a series of pericdic reports in which we de- scribe briefly the progress of the program. ORNL-3708 is an especially useful report because it glves a thorough review of the design and con- struction and supporting development work for the MGRE. It also de- scribes much of the general technology for molten-salt reactor systems. Other reports issued in this series are: ORNIL~2474 Period Ending January 31, 1958 ORNI~-2626 Period Ending October 31, 1958 ORNI-2684 Period Ending January 31, 1929 ORNI-2723 Period Ending April 30, 1959 ORNI~2799 Teriod Ending July 31, 1959 ORNI~2820 Period Ending October 321, 1959 ORNI~2973 Periods Ending January 31 and April 3G, 1260 CRNL-3014 Period Ending July 31, 1960 ORNI~3122 Period Ending February 28, 1961 ORWNL~3215 Period Ending August 31, 1961 ORNI~3282 Period Ending February 28, 1962 ORNT~3369 Period Fnding August 31, 1962 ORNI~3419 Period Ending January 31, 1963 ORNI1-352% Period Ending July 31, 1963 ORNTL-3626 Period Ending January 31, 19064 ORNI~-3708 Period Ending July 31, 1904 ORNI~3812 Period Ending February 28, 1965 ORNI-3872 Period Ending August 31, 1265 ORNI-3936 Period Ending February 28, 1966 Part 1. MOLTEN-SALT REACTOR EXPERIMENT 1. MSRE OPFRATTIONS AND ANAIYSIS P. ¥N. Haubenreich 1.1 Chronological Account of Operations and Maintenance R. H. Guymon H. C. Roller J. L. Crowley R. C. Steffy T. L. Hudson V. D. Holt P. H. Harley A. T. Krakoviak H. R. Payne B. H. Webster W. C. Ulrich C. K. McGlothlan R. Blumberg At the beginning of this report period the reactor was down to cope w1th the problem of plugging in small passages in the fuel off-gas sys- tem. Investigation showed that the material causing the trouble was a mixture of hydrocarbons, probably pump lubricating oil that had been affected by heat and radiastion in the off-gas line. With this informa- tion, a device 0o clean up the gas stream was deslgned, bullt, and in- stalled in the off~gas line upstream of the fuel pressure control valve. It consisted of a trap for particles and mist, followed by a charcoal ped for vapor adsorpticn. Also during thisg shutdown, the pressure con- trol valve and the charcoal-bed inlet valves were replaced with valves having larger trim. ‘ Power operation was resumed in mid-April, and the program of in- vestigating the performance up to full power was completed by the end of May. (This period of operation was designated run 6.) During this time, observations were made as planned on radiation heating, heat transfer, xenon poisoning, and reactor dynamics. With the exception of heat transfer, the system behavior was within the expected limits, although xenon poilsoning was somewhat less than predicted. The power escalation waz interrupted twice, once when the anomalous behavior of the control system led to a drain and again by an electrical failure in the fuel sampler, whose repair required the fuel be dralned. The approach to full power also disclosed problems in several areas: 1. After about six weeks of operation, pressure drop across the newly nstalied trap in the off-gas line had bullt up to 10 psilg, and it seemed 1t would have to be replaced. But during further operation at high power, the pressure drop went down to less than 1 psig. Temperatures indicated considerable retention of fission products in the trap but no evidence of poisoning or deterioration. o After each step up in power the pressure drop across the main char- coal bed inlets increased, hut each time this occurred, bvackblowing with helium proved effective in restoring the pressure drop to an acceptable level, ‘ 7 3. Radicactive gas from the fuel drain tanks diffused back into the helium supply lines outlside the shield, requiring & periodic purge of these lines to keep down radiation in the North Eleciric Service Area. 4. When the power reached 5 Mw, radiolytic gas began o accumulate in the thermal shield, displacing part of the water. Efforts to elimi- nate the gas by venting the shield slides and increasing the water flow were unsuccessful. 5. The water circulated through the shield contained lithium nitrite as corrosion inhibitor, and oxidation of the nitrite ion to nitrate was also observed at the higher power. 6. A discrepancy arose belween the reactor power indicated by the heat balances and by the neutron instruments (calibrated at low power ) . 1t was found later that the neutron instruments gave errconeously high readings at high power because of an 1lncrease 1n the temperature of water in the shaft around the instruments. The climax of the power escalation was reached when the coolant heat removal system (air-cooled radiator with adjustable doors, bypass damper, and two blowers) was extended to its limit, and the capability proved to be only 7.5 Mw. The coefficient for heat transfer in the pri- mary heat exchanger was also below expectations, imposing a lower limit of about 1210°F on the fuel outlet temperature at the maximum power. In late May, power operation was halted when 1t became evident that apparent inleakage of air into the reactor cell had begun to exceed the specified limit. The cell was pressurized to 20 psig for leak hunting, and no leak of any consequence was found. It was discovered that the apparent cell leak was actually nitrogen leaking into the cell from pressurized thermocouple penetrations. When this was taken into account, the measured reactor cell leakage at 20 psig was acceptably low. While the cell leak hunt was in progress, attempts were made to de- termine the source of water which had appeared in the cell atmosphere (1 to 2 gprd had begun to condense in the component-coolant pump domes, the coolest spot exposed to the cell atmosphere). 'The source was nob located because elevated temperatures in the cell prevented meaningful leak rate measurements on separate portions of the water system. Operation at full power was resumed on June 13 to Investigate the chemistry and reactivity behavior of the reactor and to increase the ex- posure of the core specimens before their removal, scheduled at 10,000 Mwhr. This was ruan 7. Fuel-salt samples, taken at a rate of three per week, showed no change in the main constituents of the salt, very low oxide (about 50 ppm), and no appreciable growth of corrosion-product chromium in the salt. The sampler-enricher was also used to expose metal wires briefly to the gas in the pump bowl, and thesge showed more fission products than expected. Reactivity transients following changes in power indicated that the 139%e roisoning was only 0.4% 6k/k at full power. The presence of circulating gas bubbles was suggested by the low xenon poilsoning, and experiments on the effect of pressure on reactivity confirmed that there was ag much as 1% by volume of bubbles in the core salt. The system re- activity showed very 1little net change connected with power operation other than that atiributed to xenon, even though the calculatsed samarium poisoning increased to zbout 0.3% 8k/k. Ancther anomalous behavior was the slow, continuous accumulation of fuel salt in the overflow tank at pump-bowl levels below that at which accumulation had heen observed before. There was some encouragement from the treated-water system. Partial deaeration of water going to the thermal~shield slides significantly re- duced the holdup of radiclytic gas, leading to plans for more effective deaeration. Decomposition of the nitrite corrosion inhibitor leveled off at an acceptable level. Power operation was interrupted for 14 hr after a shaft coupling on one of the main blowers falled. Another interruption, this one for four days, resulted when an electrical short in a component-cooling pump caused the fuel system to drain. Power operation ended on July 17, when the hub on one of the maln blowers broke up, reducing the air flow and causing pieces of hub and blades to fly into the radiazator. The coolant system was immediately drained and cocled down to inspect the radiator. After some experiments at low power, the fuel was also drained to begin the planned program of malntenance. Integrated power up to the shutdown was 7823 Mwhr. Some of the operating data are summarized in Table 1.1. The Tirst step was to flush the fuel system with Clush salt in preparation for the specimen removal. During an experiment to verify the fuel pump overflow point, flush salt got into scome of the gas lines on the pump bowl and froze there, adding another job to the shutdown. The agsembly of graphite and Hastelloy N specimens that had been in the core since September 1965 was removed for examination. These showed practically no corrosion or deterioration. The samples were sub/Jected to & program of analysis and testing with emphasis on fission product deposition. Two control-rod drives were removed and faulty position indicators repaired. Testing showed that the water leak in the reactor cell was from one of the space coolers, and it was removed and repaired. Special teols and heaters were devised, and the frozen salt was thawed from the plugged gas lines on the pump. All the work in the reactor cell was done through the maintenance shield because of radiation, which ranged from 1 to &0 r/hr over openings in the top shield. Inspection of the radiator showed only inconseguential damage due to the blower failure, but con- siderable repair of the enclosure and the door seals was necessary be- cause of heat-induced deformation and fallures. Replacement rotary ele- ments with stronger hubs and lighter blading were procured for the main blowers. - 10 Table 1.1. bBumnary oi Some MSRE Operating Data February 28, 1966 August 31, 1966 Time critical, hr 292 1775 Integrated power, Mwhr 35 7823 Fuel loop time ahove 900°F, hr Circulating helium 1916 2780 Circulating salt 2836 4691 Coolant loop time above 900°F, hr Circulating helium 1933 2020 Circulating salt 1809 5360 Heating cycles Fuel system 5 5 Coolant system 5 5 i1l cycles Fuel system 13 20 Coolant system 7 Power cycles Fuel system 6 20 Coolant system 3 17 1.2 Reactivity Balance J. R. Engel The reactivity balance, as calculated for the MSRE, is a summation of all terms which affect the system reactivity. Tt is routinely com- puted every 5 min while the reactor is critical for the purpose of re- vealing any anomalous effects that may be developing. The value of the reactivity balance in revealing unexpected effects is strongly dependent on the accurate calculation of all the known terms. Very early in the power operation it became apparent that the 13°Xe poisoning was much iess than expected. Accordingly, this term was eliminated from the re- activity balance until the xenon effect could be evaluated and an accu- rate representation programmed into the computer. (The considerable effort directed toward evaluating the xenon effect is described in Sect. 1.3.) 1In spite of this omission, considerable useful information was derived from the reactivity balances. 11 Reactivity Balances at Power Figure 1.1 shows the history of the partial reactivity balance (no 135%e correction) during the approach to full power from 1 Mw and steady operation at high power. The power history for this report period is included for reference. Xenon Polsoning. The most prominent feature of the reactivity- balance behavior ig the transients that accompany significant changes in power. The directions of the transients and their time dependence cor- respond to the expected behavior of 13536 poisoning, but they are much smaller in magnitude. We used these transients for a tentative evalu- ation of the xenon polsoning effect. Power Coefficient of Reactivity. A power coefficient of reactivity is used Lo account for the fact that the average graphite temperature is higher than that of the fuel at high power. Initially we calculated an effective temperature difference of 44°F at 10 Mw. This leads to a """ - yrtTTTE LT i i O|'* .‘,,,l:w I e, H?‘,fi% ........ Longdee oo ’ e _1 4_* ST b L. J ‘..‘ s b "»-A T""_‘L;" e, ,“ ’ le, | | f.v‘...“ Lo &L l 4 b ‘ F St LT L T LR s e 0L Jr L NETREACTIVITY | e e || PR a#,,,w ,,,,, | oal L 11 1] Lmll L mmlmLJ“ _____ tmk“ | | 72 N - i | i J 1 | D 9 i1 13 15 17 MAY, 1966 x 4* v el g T T b S i [ e b w — ! Pl L ¥ o3 0P et N_ETREACTW'T%__I___ T ] ¥ _oal = 1 o { ,,,,, ‘[J AAAAA ‘ __J, ,,‘,’"%.af L 1 , : L ol * E L 19 21 23 25 27 JUNE, 1966 JULY, 1966 Fig. 1.1l. Medified Reactivity Balance During Power Operation. 12 predicted power coefficient of —0.006(% 8k/k)/Mw when the reactor outlet temperature is held constant. Observations of control-rod posltions immediately before and after large power changes indicated a smaller power coefficient of reactivity. Careful measurements in times that were short relative to the larger reactivity effects gave a value of —0.001 * 0.001(% 8%k/k)/Mw. This value corresponds to an effective temperature difference between the graphite and Tuel of 30°F at 10 Mw. Although the contribution of the power coefficient to the net reactivity is small, all the reactivity balances were corrected to the observed value of this parameter. Reactivity Balances at low Power A somewhat less-obvious feabture of the general reactivity behavior was a gradual drift in the positive direction of balances taken at very low power when xenon poisoning was negligible. By the time the reactor was shut down in July, this drift had raised the apparent net reactivity to +0.1% Sk/k. However, the reactivity balances do not contain a term to correct for circulating bubbles, and there is some evidence (see Sect. 1.4) that the bubble fraction increased from essentially zero to about 1% during this period. Bince circulating voids have a negative effect on the reactivity, proper compensation for the bubbles would re- veal an even larger positive drift in the net reactivity. If the bubble fraction did increase by l%, the net positive shift in reactivity during the period in question was about 0.3% 8k/k. Several of the major terms in the reactivity balance are accurately known, as demonstrated by the good balances obtained in the initial periocd of low-power operation.l These terms include the temperature coefficient of reactivity, compensation for the operating 235y concentration, and control-rod poisoning. The terms which were introduced by extended oper- ation of the reactor at high power are polsoning by low-cross-section fission products, 235y burnup, and samarium polisconing. (Poisoning by T35%e has not yet been included because of the discrepancy between pre- dicted and cbserved behavior.) One possible explanation for the upward drift in reactivity is overcompensation for one or more of these negative terms. Low~-Cross-Section Fission Products. Since the yields and effective cross sections for these Tission products are reasonably well known, this poisoning term can be accurately calculated. Furthermore, at the burnup achieved so far, this term accounts for only —0.004% 8k/k. 235y Burnup. The cross sections for 2350 were adeguately treated in the MSRE calculations, as evidenced by the prediction of the initial critical concentration and the concentration coefficlent of reactivity. Therefore, compensation for the burnup of 235U, which amounted to —0.128% 8k/k when the reactor was last shut down, is probably not a major source of error. Samarium Poisoning. We estimate that the equilibrium samarium poisoning in the MSRE will be about 1% &k/k. Both **%°gym and *5'Sm con- tribute to this effect, and, since different time constaals are associ-~ ated with each isotope, the two components are calculated separately and 13 then combined to give the total reactivity effect. The caleulated sa- marium poisoning when the fuel salt was drained on July 24 was —0.364% 5k/k. | Causes of Anomaly. The apparent reactivity anomaly at low power must result from an unknown positive reactivity effect, overcompensation for negative effects, or a combination of the two. The magnitude of this anomaly appears to be at least 0.1% 8k/k, but it may be as large as 0.3% 8k/k. The uncertainty is related to the uncertainty in the circulating void fraction during operation. To date there has been no evidence, either chemical or physical, that the items which make up the negative reactivity terms are behaving differently than was assumed in the calculations. However, additicnal detalled analyses of fulure operations will e required to verify the predictions of samarium poisoning and burnup. There has also been no outside evidence of any unforeseen phenomenon that could have a positive reactivity effect of the magnitude observed. 1.3 Xenon Poisoning B. E. Prince R. J. Kedl J. R. Engel Before the MSRE was operated at power, estimates were made of the 13%%e poisoning that would be encountered.” These estimates were based on detalled models postulated for the xenon behavior in the salt, helium, and graphite in the reactor. Values for important parameters were ob- tained wherever possible from various development tests and, in the case of the circulating bubble fraction, from experiments in the MSRE. (These early tests indicated practically no bubbles.)? On this basis the | steady~-state reactivity effect due to 135%e was estimated to be —1.08% 5k/k (poison fraction = 1.44%) at 7.5 Mw. The detailed equations which describe xXenon behavior were also used to make predictions about the time dependence of this phenomerlon.4 When the reactor was operated at power, the reactivity transients that were attributed to 1355q were found to have time constants that were close to the predlicted values, bul the magnitudes of the transients were only about 0.2 to 0.4 of the predicted values. A possible explanation for the low xenon polsoning appeared when there began to be substantial, independent evidence that there were now circulating helium bubbles in the Tuel loop (see Sect. 1.4). As a result of these observations, ad- ditional studies were performed in an effort to obtain bether correlations between our predictions and experience. Two geparate studies were performed using the same basic set of equaticons but with somewhat different objectives. The first of these was an extensive parasmeter study of the steady-state 135%e polsoning. The obJjectives were (1) to see if the MSRE experience could be explained with reasonable values of the parameters and (2) to provide information 14 about the effects of various parameters that would be required for the extension of this treatment to other molten-salt-reactor concepts. The second study was aimed primarily at the time dependence of the Xenon poisoning with fewer variations in the physical parameters. The objec- tive here, in addition to providing a better insight into the behavior of xenon in the MSRE, was the development of an adequate computational model for the inclusion of xenon in the on-line calculation of the re- activity balance. Predicted Steady-State 13556 Poisoning The steady-state 13%%e model described previously2 was modified to in- clude the effects of bubbles of hellum circulating in the salt, and calcu- lations were made. The results are shown in Figs. 1.2 and 1.3. Signifi- cant parameters which were constant for t{hese plots are as follows: Reactor power level = 7.5 Mw Salt stripping efficiency = 12% Mass transfer coefficient to bulk graphite = 0.060 ft/hr Mass transfer coefficient to center-line graphite = 0.38 ft/hr Diffusion coefficient of xenon in graphite = 1 X 104 ft2/hr some of these parameters have been changed scomewhat from those reported in the last semiannual report, but this is an updating and does not change any of the conclusions. While the calculations included transfer of '?°Xe from the salt to the bubbles, they omitted any transfer of bubbles between the salt and the surface of the graphite. There are arguments to support this omission, but the effect might later prove to be significant. Figure 1.2 shows the reactivity as a function of circulating void fraction. The band shows the variation when the bubble diameter is fixed at 0.010 in. and the mass transfer coefficlent to the bubble covers a range of 1 to 4 ft/hr. It also shows the variatlon when the mass transfer coefficient is fixed at 2 ft/hr and the bubble diameter covers a range of 0.005 to 0.020 in. For this figure the bubble stripping efficiene¥ is fixed at 10%. The bubble stripping efficiency is the fraction of '?°Xe- enriched bubbles that burst in going through the pump bowl and are re- placed with pure helium bubbles. The actual value of the bubble stripping efficiency is completely unknown, and 10% was picked only because this 1is approximately the value derived from some development tests for Lhe ef- ficlency with which dissolved gas would be stripped from the salt. Flgure 1.2b shows the variation of reactivity with void fraction at various values of bubble stripping efficiency. For this plot the bubble diameter is fixed at 0.010 in. and the bubble mass transfer coefficient is fixed at 2.0 ft/hr. The conclusion that can be drawn from these figures is that circulating bubbles certainly can, and probably do, ac- count Tor the discrepancy between measured and predicted values of 135%e 15 ORNL-DWG 66-14436 -15 ] FIXED PARAMETER: FIXED. PARAMETERS: BUBBLE STRIPPING EFFICIENCY = 10% BUBBLE DIAME TER =0.010in. l BUBBLE MASS TRANSFER COEFFICIENT 3 10 A =20 ft /hr o ] - | E ———SEE TEXT FOR DESCRIPTION 5 OF BANDWIDTH g BUBBLE STRIPPING x ~05 — e N EFFICIENCY (%) T :22222222==u- — ) (@) | (&) _.28 ¢ 0.5 1.0 15 0 05 1.0 15 CIRCULATING VOQID FRACTION (%) Fig. 1.2. 13°¥e Poisoning as a Function of Cireulating Void Fraction and Other Parameters As Indlcated ORNL *PWG 66-41437 D) g ~otl— ;6 _ > BUBBLE DIAMETER = 0.010 in. ~ T g BUBBLE MASS TRANSFER COEFFICIENT = 2.0 it /e & BUBBLE STRIPPING EFFICIENCY = 0% — & Ll & Q0 - 0.001 04 CIRCULATING VOID FRACTION {vol %) Fig. 1.3. Contribution of Systems to Total 135%e Poisoning. poisoning. To prove this conclusively, one would have to have accurate knowledge of the circulating void fraction, bubble stripping efficiency, and other factors which are unavailable at the present tinme. Figure 1.3 shows the contribution of '3°Xe in each phase (salt, graphite, and bubbles) to the tobal computed *??Xe reactivity effect. 16 The wvarious parameters are fixed as indicated. This plot shows how circu- lating bubbles work to decrease the loss of reactivity to T35%e. As the void fraction is increased, most of the dissolved xenon migrates to the bubbles, dropping the dissclved xenon concentration greatly. The con- centration potential necessary for 135%e to migrate to the graphite is reduced accordingly. DNow, since 13%%e in the graphite was the largest contribution to reactivity in the "no circulating bubble" case, the over- all loss in reactivity drops also. Analysis of Transient 13550 Poisoning The equatlions given in ref. 4 were modified to include the possible presence of minute helium gas bubbles circulating with the liguid salt. The equations describing the transient concentrations, modified to in- clude mass transfer of xenon to the gas bubbles, are given below. To simplify the description, we will assume that the neutron flux is flat throughout the core. The method described in ref. 4 for correcting for the spatial dependence of the 135%e poisoning within the graphite- moderated region is directly applicable to the modified equations given below. As in ref. 4, we have used a one-dimensional model for the fuel channels in order to simplify the calculations. The differential equa- tions governing the concentrations of *2°Xe in the liquid salt, graphite, and circulating gas volume are as follows: dNi Vc 2 ® T TN (1) s v Xe c £ c £ at 7XeP'V; } AINI - (:BXe * Rs " dXe® §7"> NXe _ hvc <.N£ __RT 8 ] ) yOVL e HXE € Xe x=0 6 _figbww@wb), (2) Xe er Xe - - ¥ g 3t T e dx2 (%Xe + UXeQ) “xe 7 () The symbols used 4 . ,NI)Xe(t) b oy Nxe(t) n%e(x, ) €, 71,%e I,Xe Xe Xe Jo T 17 yi g 1 g NXe == 0( n e(x, t) dx . (5) in these equations are defined as follows: = average concentration of 35T and 1?%Xe in the tiquid fuel salt, atoms per cublc centimeter of liquid; average concentration of 13%%e in the clreculating gas phase, atoms per cubic centimeter of gas; local concentration of 135%e at position x within the graphite stringer, measured from the graphite-salt interface, atoms per cublc centimeter of graphite; average concentration of 135%e over the graphite volume assoclated with a single fuel channel, atoms per cubic centimeter of graphite; fission density in the core galt, fissions per second per cublic centimeter of ligquid; average thermal-neutron flux in reactor core, neutrons cm™? sec“l; ' fission yields of 1351 ang lBsXe; radicactive decay constants for *3°I and 135Xe, sec“l; average absorptlion cross section of 135%e for thermal neutrons in MSRE, cm?; diffusivity of xXenon in graphite, square centimeters of graphite per zecond; ratio of volume of salt in core to effective mass transfer surface area of graphite, cm; ratio of volume of graphite to mass transfer surface area of graphite, cm; ratio of volume of s=alt Within reactor core to total volume of circulating salt; mass transfer coefficient for liguid film alt the graphite-salt interface, cm/sec; mass transfer coefficient for liquid £ilm at liquid- gas bubble interface, cm/sec; effective volume fraction of helium bubbles 1n circu-- lating fliuild; bubble diameter, cm; - universal gas constant, 82.07 em? atm mole ™ (°K)“l; average temperature, °K; 18 HX = Henry's law coefficient, cm® atm mole™; e € = graphite porosity, cubic centimeters of vold per cubic centimeter of graphite; kS = effective removal constant for stripping of 135%e from liquid, sec™t; Asb = effective removal constant for stiripping of 135%e from gas bubbles, sec™t, Equations (1) and (4) are identical in form with those given in vef. 4. Fquation (2) has been modified by the addition of the last term on the right-hand side of (2) to represent the net mass transfer of 135%s from the liquid to the gas phase. Transfer of bubbles between the liquid and the surface of the graphite was not included. Egquation (3) governs the time dependence of the 135%e in this phase. The boundary conditions re- gquired in the solution of Eq. (4) are: 3 g n?(\,. — O , (6) Ox X=¥1 ONE Xe _ £ RT g ~ % 5% ”h | e L T I0pb——m—— g B —_— _— —l— —_ = e 2 = - | ! o O ® | i ] O 4 8 12 16 20 24 28 32 36 40 44 TIME AFTER INCREASE IN POWER LEVEL () Fig. 1.5, Effect of Bubble Stripping Efficiency on Transient 335¥c Reactivity. Step change in power from 0 to 7.2 Mw; volume fraction of bubbles, 0.005. The influence of the bubble stripping efficiency on the caleculated transients is shown in Fig. 1.5. Here, the separate curves represent the variation of this parameter within a range of 10 to 20% for a constant bubtble volume fraction of 0.005. The experimental data of Fig. 1.4 are also plotted in this figure. kach of the transients shown in Figs. 1.4 and 1.5 may be separated into components which correspond to the 135%e contained in the liquid, the helium bubbles, and the graphite pores. The composition of a typical transient, corresponding to a circulating gas volume Tfraction of 0.005 and a strilpping efficiency of 10%, is shown in Fig. 1.6. For the poison- 21 ORNL-DWG €6-11440 0.5 et e T I I ] ,,,,,,,,,,,,, o TOTAL ¥9Ke POISONING | W | " [ GRAPHITE (FLAT FLUX) 02 o — e s ...,......_._......_....._........_..;;7;... bttt _ T P | — T r GRAPHITE (EFFECTIVE) O [ flo e L < A7 X . <005 ® it Q ..................... D = S 0.02 <{ = 2 S 0.0F o g z ST SOOI SRS 2 o 514 BIT PRBS-DIRECT ANALYSIS 40 | o 514 BIT PRBS-INDIRECT ANALYSIS |-I- 30 L 4 .. 20 e e v & 10—yt ] A\ - e . o g _——THEORETICAL - tal O ,,,,,,, e < g & T | A - a 40 A fl'“\‘ ------------------- 20 @%3 _30 u _ e _40 — -50 _60 ........................ ( 6} -70 ! 0.001 0.002 0.005 0.01 0.02 0.05 0.1 FREQUENCY (radians /sec) BN /N Fig. 1.12., Freguency Response: Maenitude Ratio of for N BN/ N, 7.5 Mw {a) and Phase of for Ny = 7.5 Mw (b) - . O - - 5K/ K, 35 1.9 Equipment Performance Heat Transfer — C. H. Cabbard, H. B. Piper, and R. J. Kedl As the reactor power was raised, the heat-transfer capability of the alr-cooled radiator was found to be less than expected and, in fact, to 1limit the attainable heat removal to about 7.5 Mw. The overall heat- transfer coefficient of the primary heat exchanger was also below the predicted value, resulting in somewhat larger fuel-coolant temperature differences than had been planned. fter the filrst indications of low hea®t transfer, we reexamined the reactor data to determine heat-transfer coefficients as accurately as possible and to see if the coefficients varied with power level or operating time. Meanwhile, we reviewed the original design work to see if there were errors in the caleculational method or physical properties that could account for the discrepancy be- tween the predicted and observed performance. Primary Heat Exchanger. Because of the elevated temperatures and relatively small temperature differences in the primary heat exchanger, g proper accounting for thermocouple errors is essential to an accurate calculation of heat-transfer coellicient. The performance was evaluated by two procedures, which differed mainly in the method of handling thermo- couple biases. The essentisl feature of the first procedure, used routinely at the MSRE, is a statistical fit of a theoretical relation to a large nmumber of temperature measurements at different power levels. The formulation 1s such that biases in thermocouples, if constant, do not significantly af- fect the outcome and need not be evaluated. The effect of random error in thermocouple output is minimized by using many sets of data. The relation used to evaluate the overall heat transfer coeflficient, U, is where T = measured salt temperatures, =i il mass flow rates, Cp heat capacity; 36 subscripts = Iuel, ¢ = coolant salt, i = heat e¢xchanger inlet, c = neat exchanger outlet. During operation, the terms in the derivative on the left are computed and logged by the on-line computer. These values can be retrieved and a slope determined from the plot of one against the other. The other procedure, used as a check on the results of the first, ig In some respects more straightforward. A set of temperatures is meas- ured, and a value of U is calculated from the conventional heat-transfer cqualbion where U = overall heat-transfer coefficient, Q@ = heat transferred, computed from a coolant-salt heat balance, A = 279 Tt? of total tube surface area, including the return hends, based on the tube 0D, ATm = log mean temperature difference. Temperatures used in the calculation of Al and @ are obtained by adding a bias correction to each thermocouple indication. DBilases are determined by logging a complete set of thermocouple readings when the reactor is operating at a steady, very 1ow power, so0 that the salts should be iso- thermal throughout the fuel and coolant loops. The bilas for each thermo- couple is then taken to be the differeace between the average of all the thermocouples and that particular thermocouple. Heat~transfer coefficients calculated by the two methods are shown in Fig. 1.13. The continuous curve for the "derivative" method was ob- tained by Titting data points taken over two periods of time: three weeks in April and May and seven weeks in May and June. The two sets of data gave identical results. The "conventional" points were obtained on May 26. A dependence of heat-transfer coefficients on power level, as exhibited in the results of both methods, is to be expected. As the power is raised with the core outlet temperature held constant, the average tem- peratures of the salts in the heat exchanger decrease. According to con- ventional formulas for heat-transfer coefficients, The changes in salt properties (primarily specific heat and viscosity) could account for practically all the observed variation. L ~J T & © T:: S | i OTNL—DWG |66—11442 2 METHOD OF ANALYSIS (SEE TEXT): C 800 s e e CONVENTIONAL _... 2 n . . Lt DERIVATIVE, Fig. 1.13. Observed Overall o - - - b ! e ___________ Heat Transfer Coefficients in @ 700 T N T T < \\ MSRE Primary Heat Exchanger. iy T — b & BOD o e e | e = e e & ™ « |<—1 sopl-—{+—— L i L W T — | & 400 el b9 1 2 3 a 5 6 7 8 o REACTOR POWER (Mw) Design calculations had predicted an overall heat-transfer coeffi- cient at 10 Mw of 1100 Btu hr~t £t~2 (°F)~*. (This was for the straight portion of the tubes, in a cleen condition.)} This is about a factor of 2 above the coefflcients observed at temperatures approximating those used in the design. In an effort to resolve the discrepancy, the design calculations were carefully reviewed to see 1f the proper procedures were used and to check for errors in the calculations. The results of this design review indicated that the heat exchanger had been properly designed using conventional procedures and that the design should have been con- servative by about 20%. References 11 and 12 indicate that conventional heat-transfer relations are valid for molten salts, and therefore the conventional design procedures should be applicable for the MoRE neat ex- changer. The most likely explanation for the discrepancy appears Lo be erronecus valueg of salt conductivity in the deslign computations. Recent measurements of 2 salt similar in composition to the MSRE fuel salt have shown the thermal conductivity to be about one-third of the value that was believed to be correct at the time of the heat exchanger design (see Chap. 7). No recent data are yeb available on the conductivity of the coolant salt; but 1f a similar discrepancy exists, this would essentially account for the reduced performance of the heat exchanger. Radiator. The design and instrumentatlion of the air radiator pre- clude a calculation of the overall heat-transfer coefficient at all but two power levels. At most reduced power conditions the radiator doors are partially closed or the bypass damper is open; thus the effective tube area, the air flow through the radiator, and the downstream air tem- perature are unknown. Therefore the reduced performance was not obvious until the reactor was raised to full power. The overall coefficient for the radiator was evaluated with both main blowers running, the doors full open, and the bypass damper fully closed (full~power conditions). The air flow measured at these conditions was equal to the design flow rate of 200,000 cfm. Another evaluation was made with similar radiator conditions except that only one main blower was ruflnin%. The heat-transfer coefficients were 38.5 and 28.5 Btu hr-t =2 (°F)™ and the power levels were 7.4 and 5.6 Mw for these two con- ditions. The observed heat-transfer coefficients varied with the 0.575 L Q power of the air flow rate, agreecing with a theoretical exponent of 0.6. . e . _ o oo o s But the predicted coefficient at full-power conditions was 58 Btu hr 172 (°F)"t, far above the observed valuc. The radiator design was reviewed Lo determine the reason for the un- expectedly low performance. The design procedures followed conventlional practice except in two important points: the evaluation of air properties and the allowance for error in the predictions. In the calculation of the air-slde coefficient the paysical properties of air were evaluated at the tube surface temperature instead of the prescribed "film" temperature, defined ags the mean of the surface temperature and the bulk alr tempera- ture. Had the "film" temperature been used, the overall coefficient would have been lower by 14%. Use of an erroncous value for the conductivity of * the salt had little effect on the overall coefficient, since the heat- transfer resistance on the inside of the tubes is less than 5% of the total in any case. Thus, even when the air-side calculation followed the prescribed Tormula, the observed overall coefficient was still only 66% of the predicted wvalue. This is greater than the usual error in heat- transfer predictions, but 1is probably not unreasonable considering the configuration of the radiator and the very large temperature difference . between the bulk of the air and the surface of the tubes. The radiator was designed when the nominal design power for the reactor was 5 Mw. The components were designed for operation at 10 Mw to ensure sufficient capacity; and when it was later decided to call the MSRE a 10-Mw reactor, this left them with little or no allowance for uncertalnty. The actuzal tube area is only 4% more than the minimum required for 10-Mw operation according to the original calculations. Effects of lLow Heat Transier on Reactor Operation. In the main heat exchanger the reduced heat transfer causes a larvger temperature difference between the fuel and ccoolant salts Tor any given power level. The origi- nal design temperatures for 10-Mw operation were 1225 and 1175°F for the fuel salt entering and leaving the heat exchanger and 1025 and 1100°F for the coolant salt. Actually, operation at a maximum fuel temperature of - 1225°F and a minimum coolant temperature of 1025°F results in a heat- transfer ratc of 7.5 Mw. IT would be possible to operate the heat exchanger at higher power levels by increasing the heat exchanger fuel inlet temperature and/or reducing the coolant inlet temperature. How- ever, for long-term operation the heat exchanger fuel inlet temperature is limited to a maximum of 1250°F by thermal stress and stress rupture considerations, and the coolant inlet temperature is limited to a minimun temperature of 1000°F by the possibility of freezing the radiator. Op- eration at these temperature conditions would approach 10 Mw. The heat- transfer rate could also be improved by lncreasing the fuel and coolant- salt flow rates. The Tlow rates could be increased either by installing larger~diameter impellers or by increasing the pump speeds by using a higher~frequency power supply. The reduced performance of the coolant radiator, however, imposes a definite limit on the heat~removal rate from the coolant system, and there is no convenient method of increasing the heat removal. The mean tem- perature difference between the coolant salt and the bulk-air temperature is 920°F, and the coolant-zalt temperature would have to be increased 39 significantly before a gain in reactor power could be realized. The fuel- inlet temperature to the heat exchanger would then be pushed to an un- acceptable level. In summary, the maximum reactor power level is limited by the air- side heat transfer from the coolant radiator. There is also a less severe restriction at the main heat exchanger which could pe clrecumvented by op- erating the reactor system at off-design tenmperatures. The radiator heat tranzfer can be improved only by relatively extensive modifications. There will be a small increase in maximum power level during the winter months because of the lower amplent air temperatures. Main Blowers — C. H. Gabbard Aerodynamic Performance. The aecrodynamic performance of the main blowers was satisfactory. The vane angle on both blowers had veen set at 20°, as originally specified by the manufacturer for the design con- ditions. However, this vane setting did not fully load the drive motors; 80 when the maximum reactor power was Tound to be below the expected ‘ value, the vane angle was increased to 22.5° to increase the air flow and heat removal. As expected, this setting loaded the drive motors to their capacity and increased the air flow sbout 10%. The effect was to raise maximum reactor power by slightly less than 1/2 Mw. Motors. The blowers are driven by 250-hp wound-rotor induction motors that have four stages of external rotor resistance. Auvtomatic stepping switches shunt out these resistances in a timed sequence during the startup of the blower to limit the starting current of the motors. During the early stages of power operation, difficulty was periodically experienced in the start of main blower No. 1. The motor current during the starts was erratic, and sometimes the current was high enough to trip the circuilt breaker. One of the cast-iron resistance grids in the ex- ternal rotor resistance on MB-1 was found broken. A broken grid was also found in the MB-3 starting resistors, but this grid was in & less critical location in the starting sequence and its effect had not been noticed. Both the grids were weld-repaired, and no further difficulties of this type have been noticed. ’ Coupling Failure. The blowers are comnected to their respective drive motors through short floating shafts with disk-type flexible couplings on each end. On June 14, while the reactor was operating at full power, the couplings on MB-1 failed. The shaft coupling at the motor end apparently. failed first, and the coupling on the blower end was then torn apart by the resulting shaft whip. The shaft destroyed the coupling gaard and damaged the sheet-metal nose that covered the front hearing of the blower. Debris from the two couplings was scaltered throughout the fan room, and scratches on the fan vlades indicated that some of the material had gone through the blowers into the radiator duct. The coupling on the motor end of the shaft showed evidence of O~ erating in a partially failed condition for some time. The nuts holding the flexible disks were severely worn, indicating that the motor torque had been transmitted from the motor flange directly to the shaft flange 40 by the bolts rather than through the flexible disks. The initial failure of the disks was attributed to fatigue where some incorrect, flat washers had caused high stresses. The couplings on MB-1 were rebuilt, and new Tlexible shims were in- stalled in the MB-3 couplings. Operation was resumed after the radiator duct was cleaned and inspection of the hol radiator tubes showed nc dam- age Lo the tubes. Blade and Hub Failure. Power operation was brought to a premature end on July 17 by the catastrophic failure of the rotor hub and the blading of MB~1l. The outer periphery of the rotor hub disintegrated, and all the blading was destroyed. Most of the fragments were contalined in the blower housing, but numerous fragments of the cast aluminum-alloy hub and blades entered the radiator duct and some actually passed through the radiator. The reactor was taken to very low power, and the coolant was drained to determine the cause of the failure, to inspect the other blower, and to examine the radiator for possible damage. Inspection of the broken pieces of MB-1 revealed numercus "old" cracks in the blades and in the hub as evidenced by darkened or dirty areas on the fractured surfaces. One blade in particular had falled along a large "old" crack. The hub had contained short, 1-1/2 to 2 in., cir- cumferential cracks at the base of & of the 16 blade sockets, and the failure generally followed those cracks. Figure 1.14 is a piece of the hub showing the darkened areas at the base of the blade sockets. Since the failed blower had contained these old cracks, the top casing was removed from MB-3 to permit a careful inspection of its hub and blading. The Iront hub casting contained a large, continuous crack wnich extended about 35% around the circumference. There were also several of the short cracks similar to those that had been in the MB-1 hub. The blades were dye-penetrant inspected and found to be satisfactory. A re- placement blower that had not been in service also contalned some minor surface cracks in the hub. There is no general agreement on the cause of the failure. The original soundness of the hub castings is in question because of the old appearance of portions of the fracture surfaces and the presence of AHO O 84352 Fig. 1.14. Fragment of Failed Impeller Hub from Main Blower No. 1. 41 cracks in the other blower hubs. Shock forces produced during the coupling failure and vibration due to slight imbalances are suspected as contributing factors. The three blowers are being rebuilt in the manufacturer's plant. The hubs are being reinforced to relieve the bending moment at the base of the blade sockets, and the centrifugal blade loading is being reduced by substituting magnesium alloy vlades which are 35% lighter. The three complete rotor assemblies will be gilven a 30% overspeed test with dye- penetrant inspections before and after the tests. In the installation, close tolerances will be imposed on alignment and vibration, and instru- mentation will be provided to monitor vibration during operation. Radiator Enclosure — T. L. Hudson, C. H. Gabbard, and M. Richardson Operation of the reactor at power provided the ultimate tegt of the radiator enclosure, involving for the first time the operation of the radiator at the maximum capability with the doors fully open and with forced air circulation. Operation at power levels up to 1.0 Mw had veen achieved during the last report periocd. This operation had indi- cated thalt the radiator door seals had become less effective during op- eration. The loss in door-seal effectiveness continued during this re- port perlod. ' Power Level at Various Radiator Conditions. During the power es- calatlion phase of operation, the radiator conditions were adjusted to obtain preselected power levels. Therefore, the complete heat-removal characteristics of the radiator are not known because the reactor has operated at steady-state conditions at relatively few power levels. How- ever, the reactor power levels for some of the key settings are listed in Table 1.4. All intermediate power levels can be obtained by the proper adjustment of tne doors or the bypass damper. Table 1.,4. Radiator Conditicns and Reactor Power Radiator Conditions Inlet Oufilet Bypass Main Blowers Power (1) Door Door Damper Running Closed Closed Open None 0-0.05% Cpen 15 in. .Open Open 1 2.5 Cpen Cpen Open 1 4ol Open Open Closed 1 5.8 Open Open Closed 2 7.2 O.Zb iDepends on heat leakage and heater settings. Fxact value depends on ambient alr temperature. 42 - Salt Frozen in Tubes. One of the main considerations during the design of the radiator enclosure was to avold the freezing of salt in the radiator tubes. The expansion of salt during thawing was helieved capable of rupturing a tube if the thawing first occurred in the center section of a tube, so that the molten salt was confined between two fro- zen plugs. A "load scram,"” which dropped the radistor doors and stopped the main blowers when the radiator salt outlet temperature dropped be- low 900°F, was provided to prevent salt from being frozen in the radia- tor. However, salt was frozen in the radiator tubes on two occasions and was successfully melted out with no apparent damage to the radia- tor. In both cases, the freezing occurred as a result of a rod and load scram combined with a stoppage of the coolant pump. The first freezeup occurred when a defective relay caused a 'rod scram' Trom a 5.0-Mw power level. The radiator load was scrammed manually at 1000°F because of the replidly decreasing temperatures, but the coolant pump was stopped by a - low level in the pump bowl which resulted from the reduced temperature. With circulation stopped, heat losses froze some salt in 30 or more of the 120 tubes. The second freezeup occurred as a result of an electrical failure which caused a stoppage of the coolant pump and a scram of the rods and load. In both cases the coolant system was drained immediately, but the drain-tank weight Indicated that some salt had remained in the coolant systen. Recent data indicate that the volume change during the thawing of coolant salt is relatively small. This low volume change, plus the fact that the normal temperature distribution during heating of the ra- diator would cause progressive thawing of the tubes from the top to the bottom, gave confidence that the radiator could be thawed without dam- age. In the first freezing incident, the radiator was reheated slowly and all. the remaining salt was recovered in the drain tank. A pressure test was then conducted on the entire coolant system to check for rup- - tured tubes. There was no indication of leakage. After the second ) freezing incident, increased heat leakage from the radiator enclosure prevented a portion of the radiator from reaching the melting point, - and some salt remained frozen in the {ubes. However, the lowest tem- peratures were sufficlently near the melting point that the coolant system was Tilled and circulated. The radiator temperatures indicated that four or five of the tubes were blocked at first, but after several minutes of salt circulation all the tubes thawed and reached the tempera- ture of the flowing salt. After the first freezing incident, the control system was revised so that a rod scram would alsc cause a load scram. The low-bemperature set point for a load scram was increased from 900°F, which is only 60°F above the liguidus temperature, to 990°F. Other revisions were made s0 that the coolant pump would not be turned off unless absolubely neceg~ sary. A scram test was conducted from full power, and these revisions were adequate to prevent freezing of the radiator as long as the coolant pumnp remained running. 43 Damage and Deterioration. The radiator was subjected to possible damage on two occasions from mechanical failures of main blower No. 1. Pleces of stainless steel shim stock were hlown into the radiator duct when the shaft coupling failed. The radiator tubes were visually in- spected from the inlet side while salt was circulated at operating tem- perature, and no evidence of damage from the coupling failure was found. A loose sheet-metal cover from an electrical cable tray was found against the tubes and was removed. Reactor operation was resumed after the ra- diator duct was cleaned to remove the debris from the coupling. The coolant system was drained and the radiator was cooled and thoroughly Inspected after the shutdown that followed the hub failure of main blower No. 1. ©Several radiator tubes were dented, possibly by aluninun fragments from the blower, and a few small pieces of aluminum were stuck to the tubes. These pleces were easily removed, and the tubes were cleaned to remove any adhering aluminum. Visual inspection and a helium leak test at 20 psig indicated that none of the tubes had been seriously damaged by the blower fragments, and tests were run that indicated that the exposure to aluminum would not endanger the life of the radiator if the tubes were cleaned. In addition to the damage that had been caused by the blower failure, the radiator enclosure was 1n need of other repair. Heating the empty radiator to an acceptable temperature distrivution priocr to a fill had become increasingly difficult, and on the last £ill some of the tubes could not be heated above the melting point of salt until the radiator was filled and circulation started. The radiator doors had warped a little, and the seal strips had been severely distorted. The geal strips had also been torn loose in some places by the operation of the doors. Numerous sheet-metdl screws had broken or had worked loose, allowing sheet-metal cable-tray covers and sheet metal on the inside surfaces of the enclosure to come loose. There were also numerous broken electrical insulators. Although there was excessive heat leakage from the radiastor enclo- sure, this wasg mainly around the doors, and the overheating of electrical leads and thermocouple leads that had occurred previously did not recur. The repair of the radiator enclosure is in progress. Deslgn of the seal strip has been improved to reduce the thermal distortion and the overall warpage of the doors. This seal strip is segmented and free to move to allow for differential thermal expansion, and the T- bar which holds these gegments has been slotted to relieve thermal stresses which were causging the door to warp. The loose sheet metal and the loose ceramic heater elements are being held in place with welded c¢lipse rather than the sheet-metal screws. Tuel Off-Gas System — A, N. Smith The difficulties encountered with the fuel off-gas system during the initial operation of the reactor at power, the findings and con- clusions relative to the causes of thes difficulties, and the system modifications proposed to prevent the recurrence of the difficulties Al were all described ia the last progress report. The modificaitions, which consisted essentially in using valves with larger flow areas and installing a particle trap and an organic-vapor trap upstream of the pressure-control valve (PCV~522), were all completed before resuming power operation of the reactor in April 1966. Observable pressures and temperatures in the off-gas system were carefully monitored during all subsequent operations, partly to evaluate the effectiveness of the changes but primarily to identify and correct any undesirable conditions before they became unmanageable. Bome difficulties were encountered with buildup of pressure drop, but none were serious enough to force a shutdown. In addition, there was no observable loss in efficiency of the primary function of the off-gas system, the retention of gaseous fission products. Detailed analysis of the performance of some components, particularly the particle trap and the organic-vapor trap, was hampered somewhal by the scarcity of accessible pressure-meagsuring devices in the system. This was aggravated by the fact that the pressure drop across the main charcoal beds occasionally exceeded the 3-psi range of the installed (and inacces- sible during operation) measuring instrument and also by the failure of this instrument ten days before the shutdown on July 17. Line 522 Holdup Volume. When plugging occurred at several points in the off-gas system shortly after the power was first raised, it was suggested that the dependence on power might be related to radiation heating of the off-gas holdup volume in the reactor cell. Off-gas samples taken while the reactor was shut down in March with the reactor cell at different temperatures showed more hydrocarbons at nigher tem- peratures, lending support to the hypothesis that there was a reservoir of hydrocarbens in the holdup volume. t was nol practicable to clean the 68-ft-long, 4-in.-diam pipe; but the off-gas line was disconnected at the fuel pump and in the vent house, and large quantities of helium were blown through the line in the forward and reverse directions at velocities up to 20 times normal. Very little visible material was collected on filters at the ends of the line, but there were fission products, and the amount doubled when the cell was heated from 120 to 175°% . Visual cobservation showed that the head end of the holdup volume was clean except for a harely perceptible dustlike film. A thermocouple was attached to the holdup pipe near the hesad end for monitoring tem- peratures during power operation. When the power was subsedquently raised, the temperature rose from cell air btemperature (about 130°F) at zero power to about 235°F at 7.5 Mw. The rise in temperature after a setup from zero power occurred with a time constant of about 30 min, not inconsistent with bulldup of gaseous fisgion products in the line. Line 522 Filter Assembly. The filter assembly that was installed upstream of the fuel pressure control valve (PCV-522) consists of two separate units in series. First is a filter to remove partliculates and mist; next, a small charcoal bed Lo remove organic vapors. (See Compo- nent Development, Chap. 2, for a detailed description.) As a result of the experience during this period of operation, the filter part of the assembly will be replaced with a new one of the same design. The old unit will be examined in a hot cell to determine, if possible, the cause for the variaticons in pressure drop that were cbserved. 45 In the new condibion, the pressure drops across the particle trap and charcoal bed were <0.05 and 0.7 psi respectively. With these in~ stalled, the total pressure drop in the fuel off-gas sgystem was about 2.3 psi at the normal gas flow rate of 4.2 std liters/gin and PCV-522 wide open. Thus, when the reactor-system overpressure was controlled at 5 peig, a 2.7-psi pressure drop occurred at the throttling valve. Because there were no measurements of pressures at intermediate points between the pump bowl and the upstream ends of the main charcoal beds, the pressure drop across the filter assembly was known only to be below 3.4 psi. This condition prevailed until May 9, during operation of the reactor at powers up to 5 Mw. There was no indicatlon that the pressure drop across the filter assenmbly reached the detectable limit of 3.4 psi during this time. During most of the operations after May 9, PCV~522 was kept wide open, allowing the fuel overpressure to follow the total pressure drop through the other parts of the off-gas line. This mode of operation permitted the monitoring of the pressure drop across the filter assembly. For the first ten days the overall trend in the pressure drop was upward, with increases after the power was raised and decreases after it was low- ered. On May 19, with the power at 5 Mw, the pressure drop was up to 8 pei. The power was shut down to redistribute the electrical load, but the pressure drop continued on up, even though the gas fiow was reduced. On May 20, shielding was removed, and a pressure gage was temporarily attached to a tap between the particle trap and the charcoal filter to determine which was responsible for the high flow resistance. The drop measured across bthe trap was 9.9 psi, while the drop across the charcoal was only 0.5 psi. Thus the increase in resistance was due entirely to the particle trap. System pressure was reduced by venbting from the drain tanks, and the power was ralsed to determine maximum power. With the re- actor operating at 7.5 Mw the filter pressure drop decreased to about 3 psi. Then vhen the power was lowered Lo zero, the pressure drop came down over a periocd of a day to less than 1 psi. The pressure drop re- mained low until July 12, when it began to increase gradually. The in- crease continued after the shutdown, and the unit will be replaced. The particle trap was immersed in a tank of water for ccoling by natural convecticn. Thermocouples on the outside of the trap responded to changes in power and gas flow, but the maximum temperature rise was only about 25°F. It was expected that accumulation of organic material in the char- coal filter would result in progressiwve poisoning along the length of the trap. Such poisoning would shift the location of maximum fissicon product absorption and produce a shift in the temperature profile of the trap. Figure 1.15a shows two plots of the charcoal tempersture profile, one on May 10, when about 1200 Mwhr had been accumulsted, and one on July 7, when about 7000 Mwhr had been accumulated. Except for the upward shift due to the increased power level, the basic shape of the profile 1s the same for both periods, indicating that significant poisoning had not occurred during this interval. Figure 1.15b shows the effect of variations in pump-bowl pressure on the trap tempera-~ ture profile. As would be predicted, the trap temperatures, partic- ularly near the inlet, vary inversely with system pressure, because CRNL-DWG 85-11443 P FP DATE POWER PRESSURE DATE POWER PRESSURE A6--24-56 7.2 Mw 30 psig 0 5-10-66 5 Mw 4.6 psig & 7-7-66 7.2 Mw 5.0 psig ® 7766 7.2 Mw 5.0 psig 15-14-66 75 Mw 7.0 psig 140 ey e i e | - W | | | | L ] L 130—— | e L B o o \ : A& ‘ o | o 120 N\ S N 75 millivems/ hr thermal neutrons) and halfway up the access ramp (35 mr /hr RIS, , 2 millirems/hr fast neutrons, 25 millirems/hr thermal neutrons) are high during power operation. This area 1s clearly marked with radia- tion zone slgns at the entrance and halfway down the ramp. Vent House. ©OStacked concrete vlock shielding has been added peri- odically to keep dose rates low in this area. Even so, the background radiation level is ~7.0 mr/hr at full power, and the area is & condi- tional-access radiation zone. 1,10 Instrumentation and Controls J. R, Tallackson R. L. Mocre The MERE instrumentation and controls system continued to perform well. There was the normally expected reduction in both malfunctions and misoperation of instruments as instrument and operating persconnel gained experience and developed routines. While there were many design changes, most of these were improvements and additions to the system rather than corrective measures Lo the instruments and controls. A dis- appointingly large number of fauwlty commerclal relays and electrconic switches were disclosed. These faults were in the areag of both relay design and fabrication, and corrective steps have been taken. 58 Operating Experience — Process and Nuclear Instruments — C. E. Mathews, . N. Fray, R. W. Tucker, and G. H. Burger Control-System Relays. All 115 of the 48-v de-operated relays in both the safety- and control-grade circuits will be replaced because of heat damage to their Bakelite frames. This problem was discussed with the manufacturer, who stated that overheating is a common problem with all relays of this particular model if they are continuocusly ener- gized for long periods (MERE relays have operated for two years). It is a borderline conditiorn, which he says has been corrected by changing the design to reduce by 20% the total power dissipated in the operating coil and the serieg~connected dropping resistor. An order was placed for 139 relays of the latest design. Four new spare solenold coll assemblies were purchased for the spe- cial weld-sealed electric solenoid valves., Approximately 40 valves have been in service on the fuel- and coclant-salt circulating pumnp level measuring systems and the fuel sampler-enricher for two years. The first and only coil failure ocecurred recently. Pressure Transducers. A differential pressure cell used Lo obtain pressure drop in the heliwm flow through the charcoal bed shifted its range selblting. The cell has been removed, but the cause of this range shift has not yet been determined. Thermocouples. Thermocouple performance has continued to be excel- lent. Only one thermocouple failure occurred during this period. This brings the total number of failures since the start of MSRE operation to 5 out of over 1000 couples in use. Flectronic Switches. The Electra Systems switches®? used for alarm and control of temperatures in the freeze-valve system performed without malfunction during this period. This improvement in performance is at-~ tributed to modifications reported previously;l5 vo the esgtablisnment and enforcement of more rigorous test and periodic checking procedures; to the stabilization, by aging, of critical resistors in the switch mod- ules; and to a better understanding, by operating personnel, of their use in the system. A check showed that out of 109 switch set points, 83% had shifted less than 20°F over the six-month period. No swiltches with double set points were found, and it is likely that this malfunction will not reappear. Sporadic malfunctions in control loops containing a particular model current-actuated commercial electronic switch have been a source of an- noyance, This faulty behavior was, apparently, associated most frequently with ambient temperature changes obrought on by alir-conditioning failures and with excesgsive vibration. Thorough inspection, made possible by a system shutdown, revealed that out of 61 switches in service, 38 had one or more faulty internal connections. Tnese faulty connectlions, origi- nated during manufacture, were Imperfectly soldered joints or joints which had never been scldered. Nuclear Instrumentation. Water leskage via the cable into the counter-preanplifier assembly caused several failures in the wide-range 59 counting chamnnels. The cause has been diagnosed as excessgive strain and flexing of the cable, and a redesigned cable assembly will be installed. Personnel Monitoring System. The reactor building radiation and contamination warning system was again revised to correct some defi- ciencies and improve its effectiveness. Four additional beacon lights were installed. These were located in the coolant cell, blower house, diesel house, and motor-generator room. The power circuits for the lights were revised to improve reliability. The test procedures for the system were revised to include monthly tests that actuate the entire system, including the air horns, and to actuate and test the systen trouble alarm features. Data System — G. H. Burger and C. D. Martin Several new analog input signals were added to the data system, bringing the total close to the maximum capacity of 350. These were added to obtain more information about the operaticn of the off-gas system, to measure the reactor cell leak rate, and to monitor the bearing temperatures of the main blowers and the water temperature in the nuclear instrument penetration. New programs were added to supply more reactor operating informa- tion and to retrieve operating information previously stored on the magnetic tapes. The two operating information programs calculate an average of the fuel- and coolant-salt outlet temperatures and the re- actor cell leak rate. The average outlet temperature caleulations (OAFOT, OACOT) are used extensively by the operators as a guide for the operation and control of the reactor. The data-retrieval prograns were written to retrieve and process the stored information on line or at the ORNL computer center. Both programs were used many times to list and plot old data and were very useful in helping to determine the time, cause, and effect of several reactor shuldowns. These data also provided information which was used to redesign the off-gas sys- tem and resulted in changing some of the reactor cperating procedures. The retrieval program for the compuber center was written so that all stored information including calculation results could be processed and listed or plotted. The use of this program ig becoming a routine operation, and data-retrieval requests are handled by a Jjob order card. The on-line retrieval programs are more specialized and handle only certain inputs or calculation results. These programs have to be pub into the gystem as they are needed and reguire some operator time when they are executed. The analysis group is now being trained to handle rebrieval with these programs. A new general-purpose on-line program similar to the compubter cenber program is planned. Revision of operating programs ccentinued as new requirements were determined during reactor operation. The heat-balance program wasg re- vigsed and is now used to determine the true level of reactor power. The reactivity-balance program was revised but is still not entirely correct due to uncertainties in some of the parameters used in the calculations, 60 In addition to the routine and periodic collection of operating ta, the data system was used to instrument and control further re- actor dynamics bests similar to those last reported,t® Tt was also vsed Lo contyrol the temperature of a waterial surveillance test stand by a program which simulated three three-mode analog controllers. The control program compuses the surveillance specimen temperature control set point from the reactor power, fuel outlet temperature, and an off- set temperature To match the temperature profile of the material speci- mens in the reactor core. An error signal is generated by using this set polnt and the temperature of the test-stand specimens. An output signal is then generated by the computer to control the test-stand speclmen temperatures by changing the voltage to the heaters by means of compressed-air-actuated autotransformers. During the period covered by this report, the data system has be- come a virtually indispensable tool for the operators and analysts. The acceptance and confldence in this equipment which is now shared by the operators and analysts are the result of the system's ability to supply reliable and accurate information Lo assist the operators in CRNL-DWG 66-9740 [[T"oN"TIME EB SCHEDULED SHUTDOWNS SCHEDULEGC SHUTDOWNS FOR MAINTENANCE AND FOR MAINTENANCE OF MCDIFICATIONS PERIPHERAL. EQUIPMENT MSRE COMPUTER TIME AVAILABLE e —m— p TTTTTTR TR L R FF -a-‘_,,,[,:-* - T ‘k‘"“"’_h_‘fi } - 7&_LA » 100 N-57.31 % INTEGRATED »——’_ "ON" TiIME 95 i \,,‘E 80 85 7 - =1 sof- 75 70l ssr 80 55 N 501 TIME (%) 45 40F 35 301 25}~ 201 151 10 il _J 2 7 7 e o .. R ki D b - A 1 B 1522292} 5 12 19 26| 3 10 17 24 31 < 118 25|11 B 52229 6 32027 7 14 21284 11 1825 310147 24’ 1 8152229 5 121 26 MAR APRIL MAY JUNE JuLy AUG oo : CA8Bh e e 1966 — —— - —— — —,-rfi% WEEK BEGINNING Fig. 1.16. MSRE Data System Service Record from Date of Acceptance, October 1, 1965, to September 1, 19606. 61 controlling the reactor and to assist the analysts in evaluating the operation. During this period the system continued to show a steady improvement in overall reliability. Figure 1.16 shows the availability of the system on a weekly basis. OSome of the downtimes in March and April were caused by ac power difficulties as a result of installing and getting the static inverter power supply operating. ©Since April, after the inverter shakedown was completed, the system has had only one unscheduled downtime. For the 11 months the system has been in operation, the availability is in excess of 97¢, and for the past 6 months is about 98%. The system downtimes to data total 19 for 220 hr 40 min. The data system has met all the objectives for which it was orig- inally intended, and the operation is approaching routine status. Most of the programming is complete with the exception of data-retrieval pro- grams and the reactivity-balance program. It is expected that Tubture programning and other system changes will be minimum and will only re- sult from requirements generated by continued operation of the reactor. Control~System Design — A, H. Anderson, D. G. Davis, and P. G, Herndon Control Instrumentation Additions and Modifications. Further addi- tions to and modifications of the instrumentation and controls systems were made o provide additional protection, improve performance, or pro- vide more information for the operators. One hundred twenty-five re- quests for changes in the iInstrumentation and controls system (or in systems affecting instrumentation and controls) were received and re- viewed during the past report pericd. Of these, 52 requests resulted in changes in instrumentation and/or controls, 19 were canceled, 8 did not require changes in instrumentation or controls, and 25 are active requests for which design revisions are either in progress or pending. Prior to the initiation of design changes, the requests were reviewed by persons responsible for operating the reactor and for the original design., Changes in the reactor system were not made until the neces-~ sary approvals had been obtained, Some examples of these changes fol- low., fuel and Coolant Pumps. The six relays that monitor the fuel-pump motor current were replaced to prevent unnecessary shutdowns. Relay chatter was causing spuriocus operations of the reactor flux scram set- point circuits. Modifications to the existing relays and thelir current settings did not correct this condition; therefore, new relays were in- stalled. Some spurious operations occurred during thunderstorms. It ig believed that lightning-induced transients on the power distribution system caused the very fast-acting current relays to drop oub and scram the reactor. To prevent this, the existing relays in the reactor flux geram set-point circuite were replaced with time-delay relays. After the faillure of a lube oil Tlow switch interlock had shut down the coolant-salt circulating pump during a load scram, it was proposed that all protection interlocks be removed from both the coolant- and fuel-salt circulating pump control circuits. This would prevent un- necessary punp stoppages and provide additional protection against freez-~ ing of salt in the radiator. After careful congideration it was decided 62 that these interlocks should remsin in the circults, because the protec- tion they provide for the pumps oubwelghs their disadvantages. However, new operating criteria stipulate that prevention of possible freezing of salt in the radialtor is more important than protecting the punp from possible damage resulting from loss of lube ©il or cooling-water flow. To satisfy these nevw requirements, we presently plan to install a manual switerd which (after proper administrative approval) may bne operated Lo override the pump protective interlocks. Since pump shutdown may resuli from other causes, such as actlion of overcurrent Lrip in the circuit breakers, loss of TVA power, etc., additional changes in the power dis- tribution and switechgear systems will be reguired 1f maximum obtainable puping reliability is required. DManual switeh ecircuits that will over- ride all protective interlocks in this purp control circult are now belag esigned. A weld-sealed electric pressure transmitter was installed on the fuel-pump helium supply line 516 at a point hydraulically near the fuel pump. This measurement will bte compared to the punp-~-bowl cover-gas pressure to determine the pressure drop across the lube-0il static seal between the Tuel-pump shield and the impeller shaft housing. This in- formaticon should help o determine if there are periods of conditions that should favor leakage of an abnormal amount of oil into the pump bowl, Master Control Circuits. A new jumper and asscciated circultry were installed to provide a bypass around the freeze valve 111 Trozen pernissive contact in the drain-tank helium supply valve countrol cir- cuit. The Jjumper will make i, more convenient to operate through the salt-transfer frecze valves. 1t was previously necessary to freeze and thaw freeze valve 111 three times when blowing out the Lransfer lines after a Tuvel-salt transfer. To prevent radioactive gas backup into the drain-tank helium supply lines, a continuous hellum purging system, compatible with safety re- quirements, was designed and installed. Fach supply line is purged . through a capillary flcow restrictor which is sized to limit the purge flow rate to 0,07 liter/min. The capillaries are supplied from line - 519 at a peint downstream of the contalnment block valves. It was proposed that spurious operations of interlocks in the RUN mode, OPERATE mode, main blower No. 1, and main blower No. 3 control circuits that were causing unnecessary shutdowns be prevented by re- placing the existing relays in these circuits with time-delay relays. This proposal was canceled after investigations indicated that the real cause of the trouble was erratic orveration of the new 60-kva system power supply. The performance of the power supply has been improved considerably, and these circults are now operating normally. Load Control.. To satisfy established operating criteria, addi- tional control-grade circuits were designed to provide avtomatic load setback action when the reactor is in the manual load control mode. The dnstallation of these circuits is beling delayed pending a complele review of the automatic load control system. 63 Additional safety-grade circuits were installed to provide auto- matic load scram whenever the reactor control rods are scrammed. The purpose of this revision is to help prevent salt from freezing in the radiator. The design of a system to measure the vibration of the radistor cooling-air blowers and motors is now in progress. Surplus vibration instruments suitable for this application are on hand. Thermocouples. The thermocouple system did not change appreciably, although a few thermocouples were added. Twelve with in-cell type re- mote disconnects were installed on the fuel system off-gas filter and particle trap. These are read on a multipoint recorder located in the vent house. Thermocouples were also added to the new filters in off- gas line 524, to the gas holdup volume tank in line 522, and to the bearings on the main blowers. Weigh System. The problem of leaking pneumatic selector valves in the drain-tank welgh system readout was studied. Manometer readout is accomplished by selecting parbicular weigh cell channels with the se- lector valves. The valveg are composed of a stacked array of individual valves operated by cams on the operating handle shaft. ILeaks from these switches cause a slight error in the manometer reading. Efforts to stop the leaks permanently have not been successful. Leak tests on a quick digsconnect device indicate that it would be a satisfactory replacement for each valve in the switch assembly. The valves were not replaced because the problem is not severe enough at this time to Jjustify the expense, Auxiliary Systems. A pressure-reducing valve, a flowmeber, and a containment block valve were installed in the reactor cell therxmocouple nitrogen-pressurizing supply header. The normal operating pressure was reduced from 50 to 5 pslg because of the excessive leak rate into the reactor cell, The component-coolant-pump control circuits were cross~interlocked to prevent both pumps from being energized at the same time. This was accomplished previously with contacts mounted directly on the circuit breaker starter, but this arrangement would not allow one pump to op- erate normally when the breaker for the other pump is racked out for maintenance., A new design of instruments and controls is under way for a 350- gal-capacity degassing and surge tank which was added to the treated- water system to remove gases generated in the reactor thermal shield., A tank level measurement and possibly some flow control are required., An air purge system, which includes a pressure regulator, flow indi- cator, and two solenoid-operated containment block valves, is alsc re- quired. Other minor revisions and additions were as follows: 1. The beryllium monitor was relocated from the vént house to the high- bay area., 2. The range of helium flow measuring loop FE-524-B was increased. 3. The range of the differential pressure transmitter wag increased from O3 psig %o C~10 psig. Also, the indicator in this neasuring loop was replaced with a strip-chart recorder. 4. A flowmeter was installed in treated-water line 877, 5. A new instrument power distribution panel was installed to provide additional circuits required by other modifications and for future expansion. 0. Manual switches were added in the main radiator blower damper con~ trol circuit so that the dampers can be closed when the blowers are not rumning. 7. Radiator duclt blower air flow switches were added to the annunciator circuits. 8. A time delay was provided ian the high-~bay area containment pressure annunciator. References 1. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, pp. 10-12. 2. Ipbid., pp. 69-71. 3. MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, pp. 22-23. %o MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, ppr. 87-92. 5, MSR Program Semiann. Progr. Rept. Ang. 31, 1965, ORNL-3872, pp. 55-56. . R. J. Kedl, personal communicabtion, June 1966. o 7. C. H. Gabbard, Thermal-Stiress and Strain-Fatigue Analyscs of the MSRE Fuel and Coclant Pump Tanks, ORNL-TM-78 (October 1962). 8. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, pp. 12-23. i 1.3_—'2-3 . 2 9. S. J. Ball and 1. W, Kerlin, Stability Analysis of the Molten-Saltl Reactor Experiment, ORNI-TM-1070 (December 1965). 10. T. W. Kerlin and 5. J. Ball, Experimental Dynamic Analysis of the Molten-Salt Reactor BExperiment, ORNL~TM-1647 {to be published,. 11. H. W. Hoffmen and S. E. Cohen, Fused Salt Heat Transfer - Part 111; Forced Convection Heat Transfer in Circular Tubes Containing the Salt Mixture NallO,~-NaNO3-KNO3, ORNL-2433 (March 1960). 12. R. E. MacPherson and M. M. Yarogh, Development Testing Performance Evaluation of Liquid Metal and Molten Salt Heat Exchangers, ORNL- CF~60-3-164 (Mar. 17, 1960); also ANS Meeting in 1959 (Washington, DIC.)I 13. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3930, pp.29-34. 14‘. Ibid-o 3 pp - '?8—'79. 15. MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, p. 72. 16. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, p. 49. 2. COMPONENT DEVELOPMENT Dunlap Scott The development group continued to assist in the operation and test- ing of the reactor. Much of their effort consisted of helping in diag- nosing problems and in devising and installing equipment used to solve she problems. The operational performance of the equipment is covered in Chap. 1, but the description of problems and their solutions is given below. 2.1 TFreeze Valves M. Richardson Operation of the freeze valve since the addition of the modulating air controllers has been without incident. Operation of FV-103 was im- proved by deleting the hysteresis feature of the FV-103-1A module, which caused thermal cycling of the valve before the valve temperature reached equilibrium. The module FV-103-1A now operates as an off-on switch which alarms at 1000°F. 2.2 Control Rods M. Richardson The three control rods have operated without difficulty. Fiducial zero positions are listed below. Changes are caused by changes in rod length. Rod 1 Rod 2 Rod 3 Date 1.74 1.55 1.40 2/12/66 (startup) 1.73 1.43 1.49 L2466 1.78 1.57 1.52 7/14./66 Rod~drop times for 51-in. fall for rods 1 and 2 remain at < 0.8 sec. Drop time for replacement rod 3 remains between 0.9 and 0.95 sec since installation. 65 2.3 Control-Rod Drive Units M. Richardson J. BR. Tallackson The control-rod drive units were operated from installation in February until shutdown in August with the following difficulties. 1. The coarse-position synchro torqgue transmitter of the No. 3 drive Tailed. The transmitter was deenergized, and the drive unit con- tinued operation using the fine-position synchro transmitier and the po- tentiometer to indicate position. 2. The "Fill Permit" position potentiometer of the No. 1 drive began to transmit erratic readings. 'The potentiometer was Jjumpered out of the circuit, and the drive continued to operate satlsfactorily. 3. The high-temperature (200°F) switches which are mounted 6 in. from the bottom of the drive-unit housings went into alarm on No. 1 drive and remained in alarm under normal operating conditions. The switches cleared when the reactor was drained and the cell temperature was lowered. The potentiometer and synchro were replaced during the current sihut- down. FExaminatlon showed that the potentiometer had developed a region of cpen wiper contact. The synchro, mildly radiocactive, has not yet been disassembled to determine the exact cause of Tailure. Resistance measure- ments indicate a coll-to-coil short circuit in the stator windings, and the partially melted plastic end cap is positive evidence of excessive temperature. The thermostatic temperature switches® in the lower end of the drive housing are intended to indicate the approach of high-~temperature con- ditions which would damage the motor and gear box assenmbly at the uppe end of the housing. The switches are inaccessible, and it was not possible to determine absolutely if the high-temperature indication from the No. 1 drive was correct or if one or both of the switches had misoperated. Since neither of the other drive units showed a similar indication, there was room for doubt. Measurements were taken of the winding resistances of the drive motor, the fan motor, and the tachometer generator and compared to those of similar units at room temperature and in the other two rod drive units. These indicated that temperatures of the sensitive parts of the gquestionable unit were pormal. Operation was continued with the switches bypassed but with periodic checks of these winding resistances. Inspection of the No. 1 and No. 3 rod drive units in August revealed that the switches on the No. 1 unit were operating correctly, aand they showed evidence (discoloration) that they had been exposed to high tem- perature. Motors, gear boxes, and other parts of the drive units showed no evidence of overheating or mechanical difficulty. The gear cases were opened and were in excellent condition. The APL grease was soft and adhering to the worm and worit gear. The lubricant had darkened slightly, but there was no evidence of tarring or lumping of the grecase. The wire insuvlation was in good condition. 67 It is reasonable to conclude that the air stream introduced at the upper end of the No. 1 rod drive housing, while sufficient to keep the gear-box components coocled, was not large enough to prevent the develop- ment of the nigh temperature at the lower end of the housing caused by a part of the heated gas stream rising from the thimble. The temperature switches on rod drives 2 and 3 went into alarm during periods when the reactor cell was operated above 150°F. Thie indicates that the air flow to all three rod drives is marginal. The good condition of the gears, motors, and lubricant makes immediate action unnecessary, but a study will be made of methods of increasing the alr flow to the drive housings. 2.4 Radiator Doors M. Richardson Modifications® to the radiator-door-operating mechanism permitted satisfactory operational performance of the doors through the last run. However, excessive leakage of air into the radiator enclosure due to poor seals became evident after the doors had been thermally cycled several times during operation. Examination of the doors after shutdown revealed that the metal seal surfaces had been severely damaged. The metal seals, in turn, damaged the mating soft seal mounted in the face of the radiator. The metal strips had buckled between weld points, broken at some points, and been torn away completely at the top of the outlet door. Some of this damage is shown in Figs. 2.1 and 2.2. The door structure and insulation boxes were in good condition ex- cept for bowing of the 4~in. H-beam structural members. The maxinum bow, which was at the top of the outlet door, was 5/16 in. Laboratory tests were conducted on several arrangements of the metal gsurfaces. A method of thermal cycling the test sections was devised which would duplicate the distortion found in the seal strips on the doors after the previous operation. Figure 2.3 shows three of the strip ar- rangements that were tested; the strip support member (or T-bar) was the same in each case. In the case of sections 1 and 2 of Fig. 2.3, these strips were welded to the T-bar at intervals of 2 to 4 in. Because of the heat loss through the door, these strips were operated about 500°F above the temperature of the support bar, resulting in severe distortion of strips between the welds. In some places the welds had broken, per- mitting the seal strips to protrude away from the plane of the seal sur- face. A similar test was run on a third arrangement of this seal which consisted of 2-in. segments of the seal held in place at one end by a plug weld and at the other end by an overlapping tab that interlocked with the adjacent segment. This arrangement is shown as section 3 in Fig. 2.3. It was found that thermal cycling did not affect the alignment of these segments. The segments were spaced 0.031 in. along the T-bar. 68 PHOTO B5416 j ¢———SEAL STRIP SUPPORT MEMBER, SEAL STRIP TORN AWAY INSULATION BOXES BUCKLED METAL SEAL STRIP g ) SEAL STRIP —0 <«——SEAL STRIP SUPPORT (T-BAR) Fig. 2.1. Radiator Outlet Door Showing Buckled, Broken, and Torn Seal Strip (a) and Top Seal, North Side i PHOTO 854147 BROKEN GASKET RETAINER RADIATOR FACE SOFT GASKET Fig. 2.2. BSoft Seal and Retainer Outlet Side of Radiator, Showing Buckled and Broken Soft Seal Retaining Strip. 69 PHOTO 73612 .l ® SECTION 1 BROKEN PLUG WELD SECTION Fig. 2.3. BSeal Strip Tests — Radiator Door. 70 It was evident throughout all the test program that the seal-strip support member bowed about 3/16 in. along the 32-in. length of test section. It seemed reasonable to assume that bowing of this member would also occur along the solid 10-ft 4-1/2-in. length of the T-bar in the radiator door and contribute to the stressing of the door structure. The door repair work included cutting the T-bar into segments; when this was done, the bowing of the door structure was reduced from 5/16 in. to 1/8 in. Corrective measures to the seals and T-bar are presently in progress. These include: 1. Modification of the T-bar to minimize bowing by installation of ex- pansion joints in the bar. 2. 1Installation of a new hard seal which is composed of 3-1/4-in.-long overlapping links. The short links will be plug welded to the T-bar by a single weld per link with a l/32—in. expansion Jjoint between overlapping segments. 3. The existing soft-seal retainer, which is mounted on the face of the radiator, 1s being modified to permit thermal expansion of the metal retainer. 4. It is not anticipated that bowing of the T-bar will be completely eliminated. Thermal expansion of the radiator face may also con- tribute to the air leakage. An additional seal in the form of a resilient, soft seal is being proposed as backup to the existing seals. This seal would form a closure at those points along the hard seal which are not in contact with the existing soft seal. 2.5 Sampler-Enricher R. B. Gallaher During runs 6 and 7, 35 samples were isolated with the sampler-en- richer, 10 of which were 50-g oxide samples. To date 119 10-g and 20 50-g samples have been taken, and 87 enrichments made to the fuel system. In the past six months, four major maintenance Jjobs were done on the equipment. They were (1) replacement of the manipulator boots, (2) re- placement of the drive unit latch, (3) repair of open electrical circuits, and (4) recovery of a capsule which had fallen onto the gate of the op- erational valve. A brief discussion of each Job follows. Replacement of Manipulator Boots A 12-psi pressure difference which was accidentally placed across the manipulator boots caused a small puncture in the inner boot. To re- place the boots the manipulator assembly was removed from the sampler. The radiation level 3 in. from the hand was 10 r/hr as removed, but this 71 was reduced to 1 r/hr by scrubbing with soap and water. About 2 hr was required for the Job of replacing both boots. The boots had been used during the 48 sampling cycles since the operation at power was started. Replacement of the Capsule latch Just after the boots were replaced, the drive-unit motor stalled as the latch which holds the sample capsule to the cable was being retrieved through the lower bend in the transfer tube. After several tries the letch was completely withdrawn. Testing showed the latch was Jamming in the transfer tube near the top of the lower bend. The design was changed to provide additional 1/8 in. clearance between the tube wall and the lateh. The old latech and part of the cable were pulled through the access port and the removal valve with the reactor operating at less than 100 kw. The old latch was pulled into the sample transport cask to reduce the radiation level in the work area. It was disconnected from the drive unit cable and the new latch was installed. Repailr of an Cpen Electrical Circuit While operating the drive unit to be certain that the new latch would pass around the lower bend without Jamming, the cable position indicator stopped at 15 £t 2 in. from full withdrawn position, and the upper limit switeh activated. FElectrical continuity checks showed open circuits to both insert and withdraw windings of the drive-unit motor and to the upper limit switch. All three of these l€ads penetrated the containment wall through a common 8-pin receptacle. When the drive nmotor stopped, the cable extended almost to the pump bowl, preventing closing of the operational and maintenance valves. Therefore, the reactor had to be drained before maintenance work could be started. After the reactor was drained, the 15 ft 2 in. of cable was pulled out of the transfer tube without exposing the fuel system to air by using the one-hand manipulator to pull and the transport container and access port operators to hold the cable. Then the operational and main- tenance valves were closed to provide the containment barrier to allow the motor assembly to be removed for repalr. The assembly, which contains the drive-unit motor, was removed from the sampler-enricher and placed in the equipment storage cell for decon~ tamination and repair. Shadow shielding and partial decontamination were used to reduce the radiation level enough to permit direct maintenance. Partial disassembly revealed that three connector pins had burned off one 8-pin receptacle. The damaged receptacle was removed and a new one was welded in its place. All six receptacles in this location were filled with epoxy resin Lo provide additional electrical insulation and mechanical strength to the connector pins. As the drive-~unit cable was being pulled from the transfer tube it was bent in several places and needed to be straightened. The cable was = serubbed with soap and water until the radiation level was less than 5 r/hr at 3 in. Then mogt of the kinks were straightened. The remaining bends in the cable did not appear to adversely affect the operations of the drive unit when it was operated prior to reinstallation. The parts were reassembled. All electric circults were checked for continuity and for grounds, and all gas lines were Jeak checked. The assembly was then reinstalled in the sampler-enricher and the normal sampling was resumed. Recovery of a Capsule As a capsule key was being inserted into the latch, the manipulator slipped and the capsule was Jerked through the access port and fell onto the gate of the operational valve in the transfer line. A magnet was lowered into the transfer line; the mild steel key3 on the capsule was picked up by the magnet, and the capsule assembly was removed as the magnet was withdrawn. The radiation level of the magnet after this op- eration was 10 r/hr at 2 ft. These four maintenance Jobs involved handling eguipment that had been in contact with fuel salt. All work was done in contamination con- trol zones. Contamination was found outside these zones only twice, and both times it was from contaminated shoes. Apparently this type con- tamination is not readily airborne. Personnel exposures did not exceed a dose of 40 millirads/day for any one individual. Operational Valve leakage While the assembly was removed for replacement of the electrical receptacle, the upper face of the operational valve gate was cleaned as muich as possible with the valve closed. Several small, dark particles were observed on the gate prior to cleaning. After cleaning, the leak rate of helium through the top seal decreased for a while. After the capsule had peen dropped onto the gate, the leak rate suddeniy increased again, indicating that another particle had lodged in the sealing area. Contamination of the Removal Valve Seals The surface of the transport container used during the withdrawing of the cable with the manipulator became very contaminated from contact with the cable. When the transport container was remcved, part of this contamination was transferred to the removal area seal. During subsequent sampling procedures particulate contamination was spread to the top of the sampler-enricher. Since the removal aresa was cleaned using damp rags and the top of the sampler-enricher was made a contamination control zone during sampling operations, there have been no further problems with con- tamination. 73 Miscellaneous Problems During one sampling sequence as the top of the transport container was being lowered over the bottom part, which contalned a 50-g salt sample, the steel wire connecting the capsule to the key caught on the edge of the transport container top and was pulled down into the threads. When the two pieces were threaded together, the wire caused the threads to gall before the pleces sealed. The sample and the transport container were ruined. ©Since then a long plastic sieeve has been placed in the bottom of the transport container to hold the wire out of the way. The sleeve also reduces the amount of contamination transferred from the sample to the inside of the top piece of the transport container. About 4 hr is presently required to decontaminate a transport con- tainer sufficiently to be returned for use. Mild steel bottom pleces have been fabricated which will be used one time and thrown away without decontaminating. Changes to the Control Circuit Three changes were made to the control circuit. 1. The annunciator, which alarmed when capsule area pressure was greater than manipulator area pressure, was removed. A permissive light was installed to indicate when capsule area pressure wags equal to or less than manipulator area pressure. This condition is necessary before the access port can be opened but is not necessary at other times. 2. A Tuse was added to the drive-unit motor circuit to protect the motor and the electric receptacles from excessive currents. The fuse is located on one of the panelboards. 3. Voltage suppressors were placed across the two motor windings to limit any high voltage peaks during starting and stopping of the motor. 2.6 Coolant Sampler R. B. Gallaher During runs 6 and 7, ten 10-g salt samples were isolated from the coolant pump bowl. A total of 45 samples including two 50-g samples have been taken using the coolant sampler. One pin on an electrical receptacle shorted during a sampling cycle. This pin was in the circuit to the indicator light which showed when the capsule had been withdrawn 18 in. from the pump bowl. The receptacle was removed and a new one welded in place. A leak rate from the lower seal of the removal vaive increased. The valve was disassembled, cleaned, and reassembled. The valve then sealed properly. 2.7 Fuel Processing System Sampler R. B. Gallsher Design of the fuel processing system sampler has been completed ex- cept for the shielding. The sampler-enricher mockup egquipment was modi~ fied to conform with the design. All parts were received, and the re- visions tc the mockup panelboards were completed. The equipment is being ingtalled at Building 7503 whenever craft are avajlable. The tubing in- stallation is 95% complete, and the electrical and instrument work is about 50% complete. The final assembly and leak checking remain to be done. 2.8 0Off-Gas Filter Assembly A. N. Smith The off-gas filter in line 522 was originally provided to protect the very fine trim of the reactor pressure control valve, PCV 522, from becoming fouled by particulate matter. Following the analysis of plugging difficulties in February and March 1966, efforts were directed toward the design of a filter which also would trap organic materials. Primary con- siderations were (1) choice of filter medium and (2) heat dissipation. Filter Medium The organic material was presumed to exist in a range of forms ex- tending from low-molecular-welght vapors to suspensions of droplets and polymerized solids. Charcoal was selected as the best available medium for the removal of the heavier organics. Preliminary tests rveported in Chap. 7 confirmed that the charcoal would have good efficiency for removal of Cg and heavier molecules. Tt was assumed that lighter molecules would prass through the main charcoal beds without adverse effects. Heat Dissipation The hezat load at the Tfilter comes primarily from beta decay of krypton, xenon, and their daughter products. The contribution due to krypton and xenon alone is estimated at 0.1 kw/liter. The contribution due to the daughter products would depend on the assumed physical model. For purposes of the filter design, the most pessimistic viewpolnt was adopted, namely, that all daughters formed between the pump bowl and the filter remain in suspension and are transported to the filter. Since the efficlency of the charcoal trap would vary inversely with temperature, a prefilter or particle trap was added for removal of solid daughters and the entire assembly was water cooled. 75 Particle Trap The design of the particlie trap is illustrated in Fig. 2.4. Gas from the reactor pump bowl enters at the bottom of the unit through a central pipe, reverses direction, and passes in succession through two concentric cylinders of porous metal (felt metal), the first somewhat coarser than the second, and a bed of inorganic (Fiverfrax) fibers. A layer of stainless steel wool is inserted at the bottom of the unit to serve as an impingement surface for material which might be thrown out of the stream by centrifugal action. A bellows section is provided in the felt-metal zone for thermal expansion. The Fiberfrax (a poor con- ductor) is compartmented between perforated metal plates to provide for better transmission of heat to the walls of the filter housing. Three thermocouples are attached to the outside of the filter housing - one near the top, one adjacent to the lower end of the Fiberfrax section, and one adJjacent to the middle of the felt-metal section. Specific design data are as follows: Felt-Metal Section Coarse Fine Material Stainless steel Stainless steel Manufacturer®s type FM 225 FM 204 (Fuyck Metals Co., Milford, Conn.) Element size 2-5/8 in. diam X 3-5/8 in. diam X 9 in. long 9 in. Jlong Filter area 0.5 £t? 0.7 ft2 Removal rating 98% > 1.4 u 98% > 0.1 p Clean pressure drop 1-1/4 in. Hs0 at 4.2 liters/min for coarse and fine in series Fiberf{rax Section Material Carborundum Co. Fiberfrax, long staple Tivers, 5 p mean dlameter Packing density 8-1/2 to 9 1b/ft’ Total weight 189 g Geometry Nine annular compartments, 4.05 in. OD x 0.84 in. ID, with thickness of two each at 1/4 in., two each at 1/2 in., four each at 1 in., and one at 1-1/4 in. 76 ORNL-DWG 86-11444 FINE METALLIC FILTER — NICKEL SPACER (TYP)., N\ FIBERFRAX ‘ (LONG FIBER)—, N y~ THERMOCOUPLE (TYP) COARSE METALLIC FILTER—, N " STAINLESS STEEL MESH / NICKEL BAFFLE (8} — A s STAINLESS STEEL BAFFLE STAINLESS STEEL BAFFLE > Pig. 2.4. Particle Trap of Line 522 Filter Assembly. Diactyl phthalate (DOP) efficiency tests® were run on the felt metal (coarse and fine in series) and on the Fiberfrax with the following re- sults: Aerosol s . i . Efficiency Material Geometry Pype Av Particle (%) P Size ! Felt metal Coarse and fine Polydisperse 0.8 u 96.7 in series, 4 . in. dianm . Fiberfrax 4 in. diam X Monodisperse J.3 pn 99.4 - '7"]_/4 in. long, bhed density 9-1/2 b/ft> The efficiency of the assembled particle trap using the polydisperse acrosol was determined to be 99.9%. Charcoal Trap The charcocal trap consists of 13 14 of l-in. IPS stalinlesgss steel pipe, arranged in three hairpin sections of approximately equal length. The trap was loaded with 1092 g of Pittsburgh PCB charcoal to give an average bed density of 0.5 g/em?. Six thermocouples were installed at 5, 12, 51, 59, 105, and 113 in., respectively, from the bed inlet. The 77 entire charcoal trap 1s enclosed in a sealed tank through which water is circulated at a rate of 10 gym and an. inlet temperature of 65 to 70°F. The diameter of the trap piping was selected on the basis of theoretical temperature profile caleculations.? One-inch IPS was selected as the size which would permit the lowest {rap operatlng temperature consistent with an acceptable pressure drop. Flow pressure~drop tests gave the following results at the normal reactor off-gas flow of 4.2 liters/min: psi Felt material 0.04 Fiberfrax 0.004 Total particle trap 0.06 Charcoal trap 0.85 Total filter assembly Q.95 The performance of the off-gas system after the installation of this filter is described in Chap. 1. 2.9 524 Charcosl Bed A. N. Smith Diffusion of activity into the fuel pump upper off-gas line (line 524 ) resulted in high stack activity on two occasions during the report period and dictated the need for a small charcoal bed in line 524 (see Chap. 1). A small upflow of gas (about 100 cmB/min) is maintained at the fuel pump shaft ©o inhibit back-diffusion of oil vapors into the pump bowl. Iine 524 serves to transport this flow to the off-gas system. The bed consists of 9 £t of 3-in. sched 10 pipe arranged in hairpin shape and loaded with 15.8 1b of Pittsburgh PCB charcoal. Bulk density of the bed is 0.47 g/cmB. Holdup time at 100 cm3/min helium flow is estimated as 2—1/2 days for krypton or 30 days for xenon. The bed was placed in service on June 9, 1966. 2.10 Remote Maintenance R. Blumberg At the start of this reporting period, radiation levels encountered by maintenance crews were mild due to the small accumulated power-hours produced. This low radliation level permitted the use of temporary, and 78 at times casual, shielding, and in-cell manipulations were thus unhampered by the restricticns of the shielding, so that the time required for com- pletion of these early Jjobs was shortened. However, by the middle of July, when the reactor was shut down for maintenance after having produced some 7800 Mwhr, the radiation level was an important factor, adding to the difficulty, time, and expense of the work. As a consequence the portable maintenance shield had to be set up for each job, and extensive health physics precautions, especilally in the area of contamination control (housekeeping), had to be employed. As of this writing, the remote main- tenance work of the shutdown is about 60% complete. Based on the experience gained to date one may make the following observations. (1) Contamination control is not a major difficulty. We have handled several pieces of eguipment that had much wipable and trans- Terable contamination withoul spreading contamination throughout the working area. The contamination doces not become alrborne readlly. Never- theless, the practice has been to use conservative handling measures; these involve the use of plastic bag coverings, blotter paper, gas masks, ete. (2) An encouraging sign is the ease of executing some of the routine maintenance jobs. There have been many instances where removal and re- placement of in-cell components have gone very smoothly and with minimuam supervision. In general, these are items which have standard electrical and piping disconnects and supports, such as the space cocler, and wnich represent a considerable percentage of the equipment that must be main- tained remotely. (3) Thus far, almost all of the maintenance technigues or systems have been tried and have worked well. These are the portable maintenance shield, graphite sample system, vent house shielding, the remote maintenance control room, and repair techniques in the decontami- nation cell. The remote maintenance tasks where we do not yet have ex- rerience are in general associated with the handlling of the large compo- nents. The shielding provided for maintenance has been guite adequate. The following 1s a summary of the tasks performed during the Jlast period. The accumulated hours of power operation, together with an in- dication of the radiation level in the immediate area Just below the work shield, are given for each task. 1In general, the work shield was ef- fective in reducing the background radiation level above 1t to less than 5 mr/hr except while a tool penetration was partially open, and then the radiation level at the hands would be near 200 mr/hr. The maxizum ac- cumulated deosage Lo any one worker was less than 1/4 of' the maximum per- migsible level over the periocd. March 1 - 35 Mwhr Removed the flexible Jumper which connects the off-gas Jine of the pump bowl to the permanent in-cell part of line 522, in- spected the Tlange faces and inside of the piping with a bore- scope and binoculars, obtalined specimen of a residue from the piping for chemical analysis, and restored tihe system. This task was part of the effort to diagnose the restriction problem in the off-gas system. The radiation level was 200 mr/hr at floor level. 79 March 10 — 35 Mwhr Removed and replaced the reactor cell east space cooler. This operation was accomplished mostly by the craft forces. Readings were 500 mr/hr at floor level. March 12 - 35 Mwhr Installed a thermocouple on a horizontal section of off- gags line 522. A C-clamp was modified to provide the contacting force to hold the couple onto the 4-in. pipe. The thermocouple leads were connected to a spare thermocouple lead-out bhox. Radiation was 500 mr/hr at floor level. March 31 — 35 Mwhr The revised PCV 522 valve and {ilter unit descrived ahove were installed in the vent house area. Changes in the shielding and containment arrangement were done to make possible the re- mote replacement of any one of three component parts, that is, the valve, the particle trap, or the carbon bed, rather than having to remove the entire assenbly. The new installation re- quired larger maintenance shielding and stronger structural support for the shielding. April 30-May 6 — 820 Mwhr Repaired the sampler-enricher. A large subassembly of the equipment was removed to the decontamination cell, where it was cleaned and repaired using local shielding. This was the first real use of the decontamination cell, which had heen prepared for this type of work, and it was found to work quite well. There were contact readings of above 100 r/hr, but the radiation was soft enough that a little shielding reduced the field to ac- ceptable levels. May 20 — 1932 Mwhr Installed temporary instrumentation piping at the PCV 522 areca for particle trap, carbon bed, and valve pressure drop neasurements. Tubing connections were made up remotely; this was a new maintenance procedure. Readings were 100 r/hr at floor level and 10 r/hr through an open tool penetration. July 27 — 7822 Mwhr This was the start of the shutdown which has extended to the end of this report period. The reactor was shut down on July 17, and the graphite—~INOR-& surveillance samples were re- moved from the reactor core on July 28, 11 days later. The work required three days. The estimated radiation level while the sample was being transported was 1500 r/hr at 1 ft. The remote control room for the crane was used for the first time and performed satisfactorily. The tools used for the operation were contaminated to a level greater than 100 r/hr. 3. ‘4’0 80 August 510 — 7822 Mwhr Removed, repaired, and replaced the Nos. 1 and 3 control rod drives. The detalils of the repair work are described in bect. 2.3. The remote handling operation of the removal and replacement went smoothly. Radiation was 500 mr/hr at floor level. August 15 — 7822 Mwhr Removed and replaced the reactor cell west space cooler. The Jjob went routinely and was accomplished mostly by the craft forces. Radlation was 1 to 6 r/hr at floor level. August 17 — 7822 Mwhr A gas pressure reference line which became filled with a salt plug was cleared by applying pressure while heating - the line. This task was unanticipated and required some de- sign, Tabrication, and testing prior to installing a heater on a long-handled tool and safely operating it remotely. Readings were 60 r/hr at floor level and 10 r/hr through a tool penetration. References MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNI~3936, p. 54. Thid., pp. 54—56. Tohid., p. 58. E. C. Parrish and R. W. Schneider, Tests of High Efficiency Filtlers and Filter Installation at ORNL, ORNIL-3442 (May 17, 1963). Personal communication from 5. J. Ball and J. R. Engel. 3. PUMP DEVELOPMENT - P. G. Smith A. G. Grindell 3.1 MORE Pumps Molten-2alt Pump Operation in the Prototype Pump Test Facility The prototype pump was operated for 2631 hr, circulating the salt LiF-Bel'p~ZrF 4~ThE 4~UF (68,424 .6~5.0~1.1~0.9 nole %) at 1200°F., The fuel-salt pump impeller of ll~l/2 in. OD (size installed in the MSRE fuel pump) wags used. Meagurements were made of the concentration of helium bubbles in the cilrculating salt, and various tests were con- ducted on the pump-tank and the catch-basin purge gas lines,t The radiation densitometer* was used to determine the concentration of helium bubbles circulating in 1200°F salt during operation at normal pump-tank liquid level and with a fuel-pump impeller of 11-1/2 in. di- ameter running at 1170 rpm. The helium concentration in the salt was 0.1% by volume., This value contrasts with 4.6% by volume measured pre- viously?® in the prototype facility using a 13-in.-diam fuel impeller, and compares favorably to a barely detectable concentration in the MaRL fuel circuit. Tegts were performed which included analyses to debermine the hy- drocarbon content of the pump-tank and catch-~basin purge gas streams and. the composition of hydrocarbons. They were done as part of the effort to establish whether the fuel pump was the likely source of hy- drocarbons in the off-gas system of the MSRE. During these tests the Tlow of gas down the shaft annulus at constant supply pressure decreased, and this was taken ag an indication of a partial plug in the ammulus. The parbial plug increased the pressure difference acrosg the gasketed joint between the shield plug and the bearing housing from O to 5 psi at the design purge flow of 4 liters/min. Under these circumstances the off-gas from the pump tank contalined several hundred parts per millicn of hydrocarbons. This indicated that the pressure difference was forcing oil from the catch basin through the gasketed seal, down the outside of the shield plug, and into the pump tank. The partial plug could be removed temporarily by stopping the pump for a short time or by ralsing the temperature of the circulating salt from 1200 to 1250°F. When the plug was not present, the concentration of hydro- carbons in the off-gas stream from the pump tank was very low. BSamples of the plug material were obtained when the pump was dismantled. Ex- amination showed them to be salt of the composition that was being cir- culated by the pump. The mechanism by which the salt reached the shaft apnulus is being investigated. Other tests were run in which oil was injected into tThe pump bank through the sampler-enricher nozzle. Results of the analyses apnd other information about both types of tests are reported in more detail in dect. 7.5, subsection entitled "Analysis of Hellum Blanket Gas.' *Densitometer was provided and operated by V. A. McKay of the In- strumentation and Controls Division. 81 g2 Pump Rotary Element Modification The modification® that replaces Lhe gasketed seal between the shield plug and the bearing housing with a seal weld, which should eliminate the oil leakage into the pump bowl, was completed on the spare rotary elements for the MSRE fuel and coolant pumps and on the rotary element for the MK-2 fuel pump. The spare element, for the fuel pump was tested in the cold shakedown stand and installed in the pro- tobtype hot test stand for shakedown. During preheat of the system and pump, approximately 1 qt of oil passed through the lower rotary seal into the catch basin. The cause of the poor performance of the seal is being investigated, and the rotary element is being cleaned and re- assembled, Cold and hot shakedown tests of the spare rotary element for the MSRE coolant pump were completed. The hydrocarbon content in the pump tank off-gas was not more than 15 ppm, Tubrication System The lubrication pump endurance test® was continued, and the pump - has now run for 27,000 hr, circulating oil at 160°F and 70 gpm. MK-2 Fuel Pump The pump tank? is being fabricated, and, when completed, it will be installed in the prototype pump test facility. Then the MK-2 fuel punp will be tested at operating conditions gsimulating those required by the MGRE. 3.2 Cther Molten-Salt Pumps PK-P Fuel-Pump High-Temperature Endurance Test . Endurance operation3'was halted as a result of a failure of the drive motor. The pump had operated continuously for 7830 hr, circu- lating the salt TiF-BeF,-ThF 4-UF, (65-30-4-1 mole %) at 1200°F, 800 gpm, and 1650 rpm., The rotor windings of the wound rotor mobor had seized against the stator. The pump has operated for a total of 23,426 hr in four tests. Pump Containing a Molten-Salt-Lubricated Journal Bearing The gimbals support for the salt bearing4 was modified, and a new bearing and journal were fabricated. Performance of the salt bearing Wwill be investigated with an oll lubricant prior to molten-salt lubri- cation. 83 References MSR Program Semiann. Progr. Rept. Feb. 28, 75. MSR Program Semiann. Progr. Rept. Aug. 31, 65fi MSR Program Semiann. Progr. Rept. Feb. 28, Tpid., p. 76. 1966, ORNL-3936, pp. 74— 1966, ORNL-3936, p. 75. 4. INSTRUMENT DEVELOPMENT R. L. Moore G. . Burger J. W. Krewson 4.1 Temperature Scanner Performance of the tcmperature scanning systeml conttinued Lo be generally satisfactory, although some problems were experienced with the oscilloscopes and mercury switches and some system instability was noted. The problems with the oscilloscopes were only continuations of those previocusly experienced due to the age and design of the scopes. The scopes are about 12 to 15 years old and have been a continuing source of trouble. Manufacture of the scopes was disconbinued, so . 1o spare parts were available. Two new solid-state-circuit scopes were ordered and installed. The new scopes have apparently elimi- - nated that problem, although more time is required to evaluate thelir rerformance. Although the mercury switches have contlnued to give much better service than expected, some problems due to normal wear developed. Upcn ordering replacement parts for the switches, it was discovered that the switches were no longer manufactured and no spare parts existed. Some spare parts and switches were obhtained from the ORNL Reactor Divisicon, but it was apparent that a replacement switch was needed. A possible replacement was found at Union Carbide Corpora- tion, Olefins Division, in South Charleston, West Virginia. The Olefins Divigion has developed a solid-state multiplexer as a direct replacement for the mercury switch and has agreed 4o sell one to ORNL for test and evaluation. An order has bheen placed for one unit, with delivery expected about January 1967. 1In the meantime the Reactor Division will coatinue to supply mercury switch spare parts as long as they are available and - can be released for our use. - The scanner instability problems mostly resulied from operating the system outside its design range. | S ; O | 4 ! a | 3 . o | ~ o8 RODS NO 1 AND 2 SCRANM Y =z 4 - - B i '__',,,.—-»--"'"‘“ #M/M,ffib,fi,, RODS NO. 1.2, AND 3 SCRAMMED 2 b f % - -] '» - | ; | i | | i ! | ; | ‘ : | T o L : 0 5 © 15 20 TIME AFTER SCRAM (sec) (x 1073 INTEGRAL COUNT 93 ORNL DN\J B6E—11447 RODS 1 AND 2 SCRAMMED TRODS ¢,2, AND 3 QCF?,’-\MMED o,s, s EXPERIMENTAL POINTS CALCULATED BY INTEGRATION OF — — REACTOR KINETICS EQUATIONS; REACTIVITY DETERMINED FROM INDEPENDENT CALIBRATION MEASUREMEN'L Fig. 5.3. Results of Rod Drop Experiments After 27 Capsule Addi- tions. 5 {0 5 20 TIME AFTER SCRAM (sec) ORNL-DWG 66-11448 umfl( T ‘ ‘ # EXPERIMENTAL POINTS g === CALCULLATED BY INTEGRATION OF B REACTOR KINETICS EQUATIONS, 8 [ e e | ’_ % 6 l T - o S L7 ~d 72 " <1 Vv & S e i ; 7 - ROD NOA SCRAMMED AFTER 30 CAPSULE ADDITIONS. : NECATIVE REACTIVITY INSERTION,AP, DETER- MINED FROM INTEGRATION OF DIFFERENTIAL ... WORTH MEASUREMENTS. ECATIVE REACTIVITY INSERTICN,1.05 Hp. C: N:CATIVt REACTIMITY INSERTION, 095Ap 0 TIME AFT!—,H SCRAM isec) Fig. 5.4. BSensitivity of Rod Drop Experiment to Changes in Magnitude of Reactivity Insertion. ez References P. N. Haubenreich et al., MSRE Zerc Power Physics Experiments, ORNL report in preparation. S. J. Ball and R. K. Adams, MATEXP — A General Purpose Digital Computer Program for 3olving Nonlinear Ordinary DifTerential Equations by the Matrix Exponential Method, ORNI~IM report in preparation. MBR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNI-3936, pp. S2-87. Part 2. MATERTATS STUDIES 6. MSRP MATERTALS G. M, Adamson, Jr. 6.1 MSRE Materials Surveillance Testing W. H, Cook Specimens of grade CGB graphite and Hastelloy N (INOR-8) were re- moved from the core of the MSRE after 7800 Mwhr of operation as part of the materials survelllance programwl The visual appearance of the metal and graphite wag good. The external surfaces were virtually free of salt. The metal was dull gray, and the graphite appeared unchenged by the expo- sure, The Reactor Chemlstry Division was supplied with graphite speci- mens for fission product analyses from the top, middle, and bottom of thie asgembly. The results of thelr analyses are reported in Chap. 7. Other exsminations scheduled for the graphlite and metal are in progress. T™e objective at this stage of sampling was to remove one stringer (one-third) of the assembly and return the others, plus a replacement, to the reactor. However, approximately one-seventh of the assenbly iua a wone just above the midplane of the reactor was severely damaged. Graphite specimens were buckled and broken, and the tensile rods were beut. Apparently, the assembly had been locked together by frozen salt during the reactor cooldown, and the large differences in the coeffi- cients of thermal expansion of the graphite and metal created high stresses which led to the mechanical damage. The damsge in this zone occurred in all three Hastelloy N and graphite stringers, requiring that all three be replaced. The graphite and Hastelloy N in the other por- tions of the array were suitable for the intended property evaluation tests. A slightly modified replacement for the damaged reactor core gpeci- mens has been installed in the MSRE. The modifications were made to re- duce mechanical stresses in the assembly by reactor heating and cooling cyeles, The Hastelloy N tensile specimen rods in the new set are made from reactor vessel wall and head materials plus two modified alloys of Hastelloy N, one containing 0.52 wt % Ti and the other 0.43 wt % Zr. These additions were made for the purpose of reducing the effects of radiation on the alloy. In general, the molten sgalts drained well from the graphite and the mebal specimens, as shown in Figs. 6.1 and 6.2. The near absence of salt ajded the disassenbly. The overall mechanical damage 1n the reactor core specimens is shown in Fig. 6.2 and in more detail in Fig. 6.3, A region of minor damsge near the bottom of the assembly is shown in Fig, 6.3a, and the region of maximum damage is shown in Fig. 6.3b. Approximately 25% of the graphite specimens were broken in each of the three stringers. The general appearance of the graphite was the same as it was when it was machined. Note the reflection of the cross pin in the surface of the graphite in Fig. 6.3a. 27 28 R-30975 200000 000000 Fig. 6.1. Basket of Reactor Core Specimens Removed from the MSRE After 7800 Mwhr of Operation. 99 R-31011 Fig. 6.2. Overall View of the Reactor Core Specimens After Removal from Basket. SLEEVE SHOULDER FLUX MONITORS SHEATH Fig. 6.3. Details of the Mechanical Damage to the Graphite and Hastelloy N Specimens in Zones of (a) Moderate and (b) Severe Damage. The graphite specimens had been assembled so that they formed tongue- and-groove joints that were bound together by Hastelloy N straps, as shown in Figs. 6.2 and 6.3. Slip-fit sleeves had been spot-welded to the metal straps that hold the rods of tensile specimens in position. The graphite was held within the cage of tensile specimen rods by cross pinning through the shoulders of the tensile specimens (Fig. 6.3a). The total assembled length of the graphite stringers at room temperature was 62 in. At the reactor operating temperature, 1210°F, the graphite and Hastelloy N had expanded approximately 0.06 and 100 0.55 in. respectively. At the freezing temperature (850°F) of the flush salt, the Hastelloy N was approximately 0.32 in. longer than the graphite. During any heating or cooling of the assembly, the greater expansion of the Hastelloy N tensile specimen rods requires that their shoulders slide within the sleeves, which were spot-welded to the straps that bind the graphite specimens. On disassembly, it was found that most of the graphite tongue-and- groove joints had separated during the heatup and allowed salt to become entrapped in the joints. As the reactor cooled, the salt had frozen in small rectangular platelets, as shown in Fig. 6.4. These ranged from 0.022 to 0,065 in. in thickness. Their typical thickness was 0.03 in., which means that the room-temperature length of each stringer of graphite and salt was approximately 62.21 rather than the normal 62.00 in. The third item found was that some of the sleeves were locked to the shoulders of tensile specimens by frozen salt. The most strongly bonded sleeves were at the ends of the severest damage, PLATELET OF FROZEN SALT & Fig. 6.4. Typical Platelets of Frozen Salt Trapped in the Tongue- and-Groove Graphite Joints. 101 We conclude that the Hastelloy N tensile specimen rods sought to return to their original room-temperature length during the reactor cool- down, but they were obstructed by the sleeves that were attached to the rods with frozen salt and by the increase in room-temperature length of the graphite columns created by the frozen salt trapped in the joints. The graphite specimens were end loaded, and they buckled and broke in the manner of slender colums. Calculations based on 45,000 psi, the lowest value, for the yield strength of Hastelloy N, showed that the end loading would be sufficient to buckle such slender colummns of graphite, assuming that the ends were fixed and the lcads were axial. As one might expect, the thinnest specimens were broken more frequently than the thicker ones. We believe that the damage was unavoidably magnified during the re- moval of the assembly from the perforated basket. We assume that a broken piece of graphite became tilted in the region that showed the severest damage, and this tilted piece of graphite acted as a yoke to bind the assembly in the basket when approximately one-third of the assembly had been withdrawn from the basket. The additional damage was probably created by this yoke when the assembly was forcibly extracted the rest of the way out of the basket. Because many finished specimens were avalilable for use in a replace- ment assembly and time was short, we decided not to make major revisions in the design of the assembly. We did, however, make three minor modi- fications to the basic design to alleviate these problems. 1. To avoid the entrapment of salt within the tongue-and-groove joints, 0,040-in.-diam pins of Hastelloy N were used to pin the joints so that the graphite would not separate during the heatup. 2. To minimize the distances that the shoulders of the tensile rods must glide within their sleeves, the middles of the graphite stringers were pinned tc the middles of the tensile specimen rods. Since expansion is now forced to work both ways from the pinned middles, the differ- ential movements are halved. However, the maximum movement, for the sleeves and shoulders farthest from the middle, is still as much as 0.2 in. 3. To minimize friction, misalignment, and salt entrapment, the sleeves were shortened and split longitudinally. Compare those of Figs. 6.3a and 6.5, The weakest point in the modifications is that salt may freeze in some of the annular spaces between sleeves and shoulders and that the pins in the tongue-and-groove joints may shear. The graphite specimens in the modified assembly were made from grade CGB graphite that is inferior to the graphite in the first assembly in that it has more cracks. The Hastelloy N tensile specimen rods of this second assembly were made from the reactor vessel wall and head materials, heat Nos. 5085 and 5065, respectively, plus two Hastelloy N alloys modi- fied with 0.52 wt % Ti and 0.43 wt % Zr, heat Nos. 21545 and 21554 re- spectively. The modified alloys are attempts to reduce the damaging 102 Y-75011 Fig. 6.5. ©Shortened and S1it Sleeves to Provide Better Salt Drainage and to Alleviate Binding. effects of irradiation. The specimens are included in the series in order to study the effects of the alloying additions on damage by irra- diation and corrosion by molten salt. Flux monitors of 0.020-in.-diam wires of type 302 stainless steel and pure iron and nickel had been sealed under vacuum in a protective sheath of Hastelloy N as shown in Figs. 6.6 and 6.3a. In the postirra- diation examination, the stainless steel and iron wires were readily separated from the sheath, although the stainless steel showed some signs of bonding to it. The nickel wire had solid-phase-bonded throughout its length to the sheath. The new set of flux monitors installed in the MSRE with the new re- actor core specimens are of the same materials. However, €0 guard against a repetition of the solid-phase bonding, the flux monitor wires and the inside of the sheath were heat treated in air to produce thin oxide coat- ings on them. To minimize vaporization of oxide, the sheath was sealed with 1 atm of air at rocom temperature. 103 ORNL-DWG 66-11449 o .100-in- M x i _ 1, - 0.100-in-DIAM x “g-in. LONG PLUG, INOR-8 g in. 1, in] I ‘ 4 3, in. 8 TIG WELD 0.125-in-0Dx 0.020-in.-WALL INOR-8 &0 ‘l/ in i i R el 2T 777 AN — ST SISO SIS I I IO, IRON NICKEL TYPE 302 STAINLESS STEEL DETAIL OF END WELDS 0.020-in-DIAM FLUX MONITCR WIRES 0o Y% NOTE: WIRES ARE FOLDED BACK AT ONE END SCALE' ! IINCH AS ILLUSTRATED ABOVE TO FACILITATE THEIR IDENTIFICATION. THE BENT ENDS ARE AT THE TOP OF THE SHEATH. PINCH CLAMP MARK (PINCH CLAMP HELD VACUUM WITHIN THE SHEATH DURING THE SEALING OF THE FINAL END WELD.) /SHEATH &= _'I:O ToP O [T = 3O ‘1/ ]«—13/ —J<— 60", * o 1y e —T 8 8 a et 62 — ¥ THE FLUX MONITORING WIRES EXTEND ALONG THIS SPACE OR DIMENSION. Fig. 6.6. MSRE Flux Monitoring Wires and Protective Sheath. 6.2 Evaluation of Possible MSRE Radiator Tubing Contamination with Aluminum D. A, Canonico D, M. Haseltine On Sunday, July 17, a cast Al-5 wt % Zn alloy blower failed at the MSRE site, throwing small shrapnel across the Hastelloy N tubing of the salt-to-air radiator. The failure occurred at about 10:00, while the radiator was operating at 1070°F. The protective doors were closed while the salt was still circulating; the temperature then rose to approxi- mately 1200°F. This temperature was maintained until 14:30, at which 104 time the salt was drained and the heat was turned off. At 19:00, the temperature had decreased to 250°F. The doors were opened and the tem- perature dropped further to 150°F at 20:00. A metallurgical investigation was conducted to determine the effect of the aluminum-zinc alloy on the radiator tubing. Since the actual tubing could not be removed and sectioned metallographically, only spec- imens simulating the exposure could be prepared and evaluated. Small segments of the failed blower were tested in compatibility experiments with representative samples of the tubing. Laboratory experiments were conducted to determine the influence of temperature and time upon the Hastelloy-N—aluminum-alloy interaction. Tem- peratures of 1150, 1175, 1190, 1200, 1225, and 1250°F were investigated, Y-74029 1250°F INCH 15 MINUTES AT TEMPERATURE Fig., 6.7. Hastelloy N—Aluminum Alloy Compatibility Test Specimens., 105 and times at temperature ranged up to 5 hr. Figure 6.7 shows the samples after being held at temperature for 15 min. Figure 6.8 shows the samples with the aluminum removed from the tubing. The refractory aluminum oxide coating had contained the aluminum, even in the molten state, and inter- getion did net cecur. The rounded ¢orners ipdicgte Lhat a lLiguld phase was present at 1175°F and above; metallographic examination verified this. When the oxide skin was broken from mechanical abrasion, shock, or other reasons, wetting occurred. Figures 6.9a and 6.9 show the appear- ance of a specimen where an angular drop of aluminum alloy had broken through its skin and become sattached to the Hastelloy N tube. The spec- imen was held for 5 hr at 1200°F. A metallographic section through the cone and tube revealed the moderate interaction shown in Fig. 6.9c. At- tack to a depth of approximately 0.010 in. had occurred. Since these experiments indicated some possibility of penetration of aluminum into the MSRE radiator tubes, they were inspected carefully for both dents and attached shrapnel, The difficulty of this task can be shown by Fig. 6.10, a photograph of the face of the radiator. The close-packed spacing of the tubes made the use of mirrors and elaborate lighting necessary. It is estimated that 80% of the surface area of the tubing was examined. Y-74028 1200°F o2 0°F g TR Bk IOk 0 1 12B0°F INCH 15 MINUTES AT TEMPERATURE Fig. 6.8. Hastelloy N—Aluminum Alloy Compabibility Test Specimens. 106 Y-74022 (e} Fig. 6.9. (a) Specimen Where an Angular Drop of Aluminum Had Broken Through Its Skin and Become Attached to a Hastelloy-N Tube. (b) Closeup of drop (10X); (c) metallographic section through cone and tube (50X). As polished. Eight smears of aluminum were found on the air-inlet side of the radiator, none on the outlet side. No cone-shaped deposits of aluminum were discovered, although one small angular piece was found wedged be- tween a lower tube and a heater. Three dents were noticed in the outer 10% HOTO 84553 Fig. 6.10. TFace of MSRE Radiator, Showing Close Packing of Tubes. tubing on the air-inlet side, but no aluminum was detected 1in these areas. It is thought that the dents resulted from small rocks, etc., which the blowers had drawn in at an earlier date. The tubes on which aluminum was found were marked and subseguently cleaned. Stainless steel wool was used to remove most of the aluminum from a deposit area; then emery paper was employed. The radiator was then brushed thoroughly to dislodge any pileces of aluminum that might have been missed during inspection. As a result of this investigation and cleanup procedure, it appears that the radiator system is satisfactory for further operation. Actual adherence of the aluminum to the tubing was observed in only a few cases. Inspection and cleaning of accessible Tubes assured that aluminum was re- moved from those most likely to be contaminated. If intimate contact occurred in scattered areas, interaction dces not appear to be rapid. Longer time effects are being studied by visual and metallographic analysis of specimens from tests of duration up to 1000 hr. 108 6.3 Evaluation of Graphite W. He Cook Work has continued on the evaluwation of anisotroplc and isotropic grades of graphite as potentilial materials for molten-salt breeder re- actors.? The MSRE graphite properties are being used as a basis for comparison. The needle-coke, anisotropic graphite has the more desir- able properties for excluding moiten salts and gases but is less stable under irradiation. Some isotropic materials show promise, but many of the isotropic grades of graphite tend to be too amorphous for nuclear use. Isotropic grades of graphite are new, and past development has not been directed toward the production of material having the low gas - permeabilities and small pore entrance diameters required of graphite for molten-salt breeder reactors. : Table 6.1 is a comparison of the physical properties of the various grades of graphite. The specific resistances of these indicate that grades - CGB, CGB-LB, 1425-64-1, and H-315 are fairly well graphitized; the rest are too amorphous. For isotropic graphite, we desire specific resis- tance less than 900 microhme em? cm™+. The gas permeabilities are all very high. We are seeking permeabilities approaching 1077 cma/sec. The pore entrance diameters are important in that (1) they must be small enough to prevent penetration by the nonwetiting fluoride salts and (2) they must be grouped to minimize the number of impregnations necessary to fabricate the base stock into a low-permeability graphite. For the latter, it has been reported that pore entrance diameters should be less than 1 u and be concentrated in a narrow range.3’4 The grades of graphite shown in Fig. 6.11 are finished grades rather than base stocks, but only grades CGB, CGB-LB, and EP1924-1 appear amenable to improvement by impregnation treatments. The third grade had been al- ready ruled out because of its amorphous nature. The standard salt screening test was made in which 0.500-in.-diam - by 1.500-in.-long specimens were exposed for 100 hr to molten salt at 1300°F under a 150-psig pressure. The results are shown in Table 6.1. The MSRE grades CGB and CGB-LB are included for reference purposes. Of the new grades, only grade 1425-64-1 showed a pore structure approaching the quality of the MSRE graphite. Work is in progress to obtain additional anisotropic and isotropic grades of graphite with properties superior to those of the MSRE graphite and approaching those required for the molten-salt breeder reactors. Table 6,1. Ccmparison of Varicus Physical Properties of Anisotropic Needle-Coke and Isotropic Grades of Craphite Specific Resistance Bulk Volunme Cross~-Section Buik Helium Accesgible Surface . 2 - Permeability of Graphite ; . . \ . microhms cm® cxm A - Grade Shape Dimensions Type Density Density Poros;ty Area to Helium Pilled (in.) (g/cn?) (gfem®) (ch )2 (2 /g )8 l!b e (cre? /fsec) Wi%h)galt - % X 104 CGB Bar 2 X2 Needle-coke 1.86 2.03 12.7 0.462 610 1200 3 0.2 CGR-LB Bar 1X 1.6 Needle-coke 1.86 2.12 12.4 595 1100 0.4 (c)wéc:"/— 0.8 1425-64-1 Pipe 3.6 0D X 2.5 ID Neecdle-coke 1.82 1.91 9.2 0.171 690 1215 4500 0.5 H-315A Pipe 4.7 00D X 3.5 1@ Isobropic 1.83 2.04 11.7 0,165 820 980 89 7.2 EP-1924-1 Cylinder 4.1 diam Isotropic 1.80 2.14 5.8 0.332 1385 2600 12.5 2020 Block 4.0 X 4.1 Isotropic 1.71 2.02 17.2 0.316 1760 1660 240G 14.6 HCTE-Y12 Cylinder 7.3 diam Isotropic 1.88 1.95 loste 0,046 13C0 1810 9460 4.0 She measurements were made on 0.250-in,-diam X 1.C00-in.-long specimens. Measured in the direction of the ap axes for needle-coke, anisctropic grapnite; no specific directions indicated for isotropic graphite ex- cept that measurements were mutually perpendicular. Measured in the direction perpendicular to the ag axes for needle-coke, anisovropic grapbite; no specific direction indicated for isotropic graphate except that two measurcments were made that were mutuaily perpendicular, Evacuated specimens, 0.500 in, diam X 1.500 in. long, exposed for 100 ar to molten salt at 1300°F and a pressure of 150 psig; a standsrd geresning Legt, Tmoregnation was not uniform; the valunes in parentheses indicate the range observed. 60T 110 ORNL-DWG 66~ H450 5o e L el e L] Z ”CGB—LB | ) _:"_ 4T‘* I l:] NJ 28 - LT = (CGB ™ i d ——— .- - . Qo 10[_; -1 T ][ : - i’:,: s ! | 2 e B g 2020 i | v o b =T - L 5} —_— _.-.1””7 cmiaaae 20— — — — — — - . sl = EPI924-1 | E_:J 0 el E {eh """" r . ‘ | | | 6 — — — — 44#4‘7 — 1425-64-19 ] ‘ o T T ] . , 209 8.0 1.0 030 003 coz ot PORE ENTRANCE DIAMETER () Fig,., 6.11. Comparison of the Distributions cof the Pore Entrance Diameters Tor Various Grades of Graphite. 6.4 The Internal Stress Problem in Graphite Moderator Blocks C. R. Kennedy One of the major conceras in the use of graphite-moderated reactors is the stress generation caused by differential growth. The stress gen- eration is moderated considerably and generally maintained at safe levels by the ability of the graphite to ereep under drradiation. The problem resolves itself {0 a determination of the balance hetween the differential growth rate that produces stress and the creep that reduces stress. Al- though the creep-~rate coefficient may be such that the gtress level does not exceed the fracture strength, a second concern is the ability of the graphite toc absorb the creep strain indefinltely without failure. Cer- tainly, a fallure criterion based upon the fracture stress of the mate-~ rial is accurate., Therefore, it is of great importance Lo know both the restrained growbh rate and the creep coefficient for the material under the operabing conditions. Our purpose has been {o determine the general creep behavior of graphite under irradiation. Creep experlments were performed at 700 and 1000°C for comparison with the previous data’ obtained at lower temperatures. The high-tem- perature experiments weve again similar to the previous experiments in 111 ORNL-DWG 66~11451 2 ‘ : " SINGLE SPECIMEN } / -4 (vt ) 4 2 r— — Fig. 6.12. Effect of Temper- 2 s ] ature on the Creep Coefficient of ¢ f Graphite Under Irradiation. b T Ll o 2 1 [ L . i . o i ‘ o 1027 G a v Sl v AGOT - w -m%_------k—f—# — © EGCR-TYPE AGOT— S iifl‘“ VALUES | 7771 a coB o 5 ,,,,,,,,,,,,,,,,, d s JA 0 200 400 600 80C 1000 1200 1400 TEMPERATURE (°C}) that cantilevered parabolic-beam specimens were used. The main differ- ence was in the uge of four-zone furnaces to obtaln the desired temper- atures. The number of specimens was reduced from nine to six because of the space requirements of the furnsces. Results from all creep experiments performed from 150 to 1000°C do demonstrate a generalized creep behavior, as shown in Fig. 6.12. This type of behavior strongly supports a Cottrell model for irradiation creep, which allows extrapolation of these data to most reactor grades of graphite. The Cottrell model for irradiation creep, as given by Anderson and Bishop,6 is K = Ajfo_ (1) y where K = creep coefficient, A = accommodation factor, ¥ = ghear rate due to anisotropic growth (GC *'Ga)’ 0&_: yvield strength of the crystallites, GC and Ga are the growth rates in the ¢ and a directions. The shear rate ¥ of graphite can be derived from measurements made on pyrolytic graphites through estimates of polycrystalline graphite growbh rates. The accommodation factor A primarily reflects the degree of accommodation by microecracks and, in general, the vold volume in the graphite. It is not necessarily constant and is expected to vary with temperature and neutron exposure. The variation of A with temperature should not be very large; however, A should exhibit a rather significant increase as microcrack closure occurs. The closure of microcracks re- gquires a dose of approximately 1022 neutrons/émz; thus the value of A will be essentially constant for at least half this dose. The value of 112 the yield strength or flow stress, 0, of the crystallites will undoubt- edly vary under Iirradiation and with irradiation temperature. The value of 0y, like that of A, will not vary with temperature, however, as sig- nificantly as ¥, the shear rate. The creep-rate coefficient should therefore strongly reflect the variation of ¥ with temperature. This is demonstrated in Fig. 6.12, where the creep~rate coefficient exhibits a minimum around 350°C. The creep-rate coefficient K is not exaclly proportional to the shear rate ¥ because of the variations of A and Oy. The value of A would be es- sentially independent of the graphite grade: thus the effect of Gy on the creep-rate coefficient can be demonstrated by a comparison of var- ious grades at the same temperature, This was demonstrated previouslys by comparing the modulus of elasticity of six grades of graphite to their creep-rate coefficients at 400°C. Although this correlation strongly supports £q. (l), or the Cottrell model, for the crecp of graphite, one glaring discrepancy in the correlation is the test result at 1000°C for the CGB grade of graphite. It should be noted that this particular ma- terial demonstrated a one-~third decrease in its modulus of elasticity, which is also unlike the behavior of previously tested grades. This particular test result requires confirming data before specific conclu- sions can be reached. One should recognize that the Cottrell model for creep does not sug- gest an actual mechanism by which the graphite deforms plastically. It does describe the creep as a process of continuous yilelding, which is not thermally activated as the term creep generally implies. Also, the stress-induced jmbalance of the internal straining occurring in all poly- crystalline graphites due to anisotropic growih. This implies that the limit of creep deformation can be as large as the internal straing that must occur in polycrystalline graphites irradiated without stress. Poly- crystalline graphite can accommodate 16% shear strain and polycrystalline carbons 160% shear strain’ without a loss of mechanical integrity. The obvious conclusion is that as long as the stress acting on the graphite does not exceed the fracture stress, the graphite will continue to ab- sorb the creep deformation without loss of mechanical integrity. 6.5 Brazing of Graphite J. M. Jones We J. Werner The jolning of graphite to structural metals such as Hastelloy N is of prime interest in advanced molten-salt reactor concepts. Studies are under way Lo develop methods for joining large graphite pipes to Hastelloy N tube sheets. OSuch joints will be needed for test assemblies and for the reactor core. The current studies have two primary objectives: (1) to develop a corrosicn-resistant brazing alloy for graphite which does not suffer from 113 the transmutation problem associated with the gold-contalning alloys and \2) to develop means for making transition joints with one or more ma- terials having expansion coefficients intermediate between those of the graphite and the Hastelloy N. Work was focused on the evaluation of an experimental precious-metal- base alloy having the composition 60 Pd—35 Ni~5 Cr (wt %). This glloy appears to have some application for Jjolning graphite to the refractory- metal portion of a transition piece. Unfortunately, it exhibits rela- tively poor flowability on high-density graphite (Fig. 6.13a); however, its marginal behavior 1s enhanced by preplacing it as foil in the Jjoint (Fig. 6.13b). Photomicrographs of such a joint at low and at high mag- nification are shown in Figs. 6.14a and 6.14b. The needle-like carbides of Fig. 6.1l4a are clearly evident in Fig. 6.14b. The extensive diffu- sion zone along the molybdenum—razing-alloy interface in Fig. 6.l4a suvggests that considerable molybdenum was taken into solution in the brazing alloy. A series of graphite-to-molybdenum Jjoints brazed with this palladium- nickel-chromium alloy preplaced in the joint were thermally cycled ten times between 200 and 700°C. Metallographic investigation indicated that no deterioration had oceurred. omall arc-melted buttons of other alloys in this ternary system, but with higher chromium contents, are being prepared in an effort to improve wetting of graphite without reducing the overall joint proper- ties. Alloys containing other corrosion-resistant carbide formers, that is, nicbium and molybdenum, are also being prepared. ITwo graphite-to-molybdenum-to-Hastelloy-N transition joints were successfully brazed using the tapered joint design reported previously.8 The 1-1/4-in.-0D by 3/1l6-in.-wall graphite tubing was joined to the mo- lybdenum with the 60 PA—35 Ni—5 Cr (wt %) alloy preplaced in the Jjoint; copper was used to braze the molybdenum to the Hastelloy N. Visual ex- amination revealed no cracks, and metallographic examination will be used to further evaluste the Jolits. Y-75419 denum Jointe Brazed with 60 Pd—35 0 1 Ni—-5 Cr (wt %) at 1250°C in Vacuum. Lok o Y-75420 1l . 0385 INCHES ra \OO)( T | Fig. 6.14. (a) Graphite-to-Molybdenum Joint Made with 60 Pd—35 Ni—5 Cr Alloy Preplaced in Joint. Good bonding is evident. Etch: 10% oxalic acids (b) View of area enclosed in (a). The large carbide needles are evident. Etch: 10% oxalic acid. 115 6.6 Corrosion of Graphite-to-Metal Brazed Joints W. H, Cock As was discussed in the previous section, some success has been obtained in joining small pieces of graphite to metal by brazing the graphite to molybdenum and, in turn, brazing the molybdenum to Has- telloy N.%'L0 TIn addition to making a good joint, the braze must be resistant to corrosion by molten fluoride salts., Corrcsion of the brazing alloy is being investigated by exposing Jjoints of grade CGB graphite bragzed toc molybdenum to static LiF-BelF,-Z2r¥,-ThF,-UF, salts for 100, 1000, 5000, 10,000, and 20,000 hr at 1300°F in Hastelloy N capsules. The brazing alloy, 35 Ni—60 Pd—5 Cr (wt %), was not at- tacked by the salt during exposures for as long as 5000 hr. However, layers of metallic-like crystals appeared preferentially on the braz- ing alloy in the 1000- and 5000-hr tests. The 10,000- and 20,000-hr exposures are still in progress. The method of fabrication of test specimens is shown in Fig. 6.15. The lateral surfaces of the machined pieces were polished prior to cut- ting the pieces into 0.,62-in.-long specimens. The microstructures of the left and right edges of the sides (the sides are perpendicular to the plane shown) of an untested braze in Fig. 6.16a show that this pro- duces straight, smooth sides suitable for references for determining the ORNL-DWG 66-11452 CGB GRAPHITE ALL DIMENSIONS ARE IN INCHES. Fig. 6.15. Grade CGB Graphite Brazed to Molybdenum with 35 Ni—60 Pd=5 Cr (wt %). (a) As brazed; (b) as machined into three test specimens. The lateral surfaces are polished. 116 GRAPHITE s < 0.030 INCH BRAZING ALLOY (a) O hr (5) 100 hr (c) 1000 hr (d) 5000 hr Fig. 6.16. Microstructures of the 35 Ni—65 Pd—5 Cr (wt %) Brazing Alloy Used to Join Grade CGB Graphite to Molybdenum. (a) As brazed and (b, ¢, d) After Exposures to LiF-BeF,-ZrF,-ThF,-UF,; (70-23.6-5-1-0.4 mole %) at 1300°F. The graphite was tilted during the brazing which pro- duced the different thicknesses of braze in the sets of photomicrographs that show both edges of a specimen. Etch: 10% oxalic acid. 100x. Y-75395 7] ]| HON| LCOC 53 Fig. 6.17. Microstructure of a Metallic-Like Deposit on the Graphite of the Graphite-Molybdenum Joint Exposed for 5000 hr to LiF-BeF,-ZrF, - Th.FA"UFé_ at lBOOoF . 117 extent of attack. Tor the metallographic examination, this specimen and all others were cut into halves perpendicular to their 0.62-in. dimension, and these cut surfaces were polished and photographed. This technigque was used to compare the control specimen with those tested for 100, 1000, and 5000 hr (Fig. 6.16). There is no microscopic evi- dence of attack on the braze. Deposits approximately 0.5 and 1 mil thick are evident on the sides of the brazes exposed for 10C0 and 5000 hr respectively. These layers are crystalline and metallic 1in appear- ance. They have not been identified. The chemical analyses of the salt from the 1000-hr test did not detect any corrosion products; the analyses for the 5000-hr test are in progress. Interpretation of the 5000-hr test is further complicated by a thin metallic-like deposit on the graphite. It is unlike the layers on the braze alloy in that it does not exhibit crystalline faces. The deposit was heaviest on the slde of the graphite parallel with the brazed Jjoint. Its microstructure is shown in Fig. 6.17. The identification of this layer 1s being sought. It is suspected that this may be CriCy. In previous tests that involved only salt, Hastelloy N, and graphite, there were no deposits on the graphite. The longer-term tests appear necessary to explain the cause of these deposits. 6.7 Welding Development of Hastelloy N H. E. McCoy D. A, Canonico Titanium and zirconium additions to Hastelloy N appear to improve the resistance of the alloy to high-temperature embrittlement in an ir- radiation field. We have made welds in several experimental heats to evaluate the influence of these alloy additions on the weldability. Welds have been made in 1/2-in. plates of Ni—12 Mo—7 Cr-0.05 ¢ (wt %) with titanium contents of 0,15, 0.27, 0.33, 0.45, and 0.55 wt %, These welds appear to he sound, and test specimens have been made for evaluation. Alloys containing Ni-12 Mo—7 Cr—0.05 C (wt %) and either 0.06 or 0.43 wt % zirconium have also been studied. The alloy containing 0.06 wt % zirconium exhibited extensive weld cracking. It was impossible to get a complete pass in the alloy containing 0.43% zirconium without a crack propagating the entire length of the pass. Some attempts have been made to weld these heats with dissimilar weld metal. Effect of Irradiation on the Mechanical Properties of Hastelloy N — H. E. McCoy The foremost objective of this effort has been to determine the in- fluence of neutron irradiation on the mechanical properties of the Has- telloy N used in constructing the MSRE. We have utilized postirradiation tensile tests, in-reactor creep-rupture tests, and postirradiation creep- rupture tests to evaluate the performance of several representative heats of material used in the MSRE. Portions of this study have been reported : 11-14 . . o . previously and have been used in setting a minimum safe operating ORNL-DWG 66-5779R I NH I 11T 70 60 ~ { UNIRRADlAT ED lrl o o / STRESS (1000 psi) W S Qo O ! ! ( . I b ""_ o 4 - /¢ (%] o | + : — 1 20 L <1> 6X1O n \H-: | \\1: i | i 1-51\3‘— ! H 2.4%%' ; 10 ; | | (\“ | 0 L 10° 2 5 0 2 5 102 2 5 10° 2 5 40° RUPTURE LIFE (hr) Fig. 6.18. Creep-Rupture Properties of Hastelloy N at 650°C, Heat 5065. Numbers indicate fracture elongations. life for the MSRE. One of the most interesting findings in this study is illustrated by the data in Fig. 6.18. Although the rupture life and ductility are both reduced by irradiation, the data indicate that there may be a stress below which essentislly no irradiation damage occurs. One of the current theories of high-temperature irradiation damage predicts this type of behavior, since some minimum stress is requlred to allow the bubbles of transmuted helium to grow without limit. 15 The existence of a stress below which one could design without concern for gross reductions in ductility due to neutron damage is of extreme im- portance and needs to be studied further. The surveillance specimens were removed from the core of the MSRE after 7800 Mwhr with an effective integrated dose of about 7 X 10%°, So far as neutron dose is concerned, these specimens should be repre- sentative of the pressure vessel after the reactor has operated about 150,000 Mwhr (eguivalent to 15,000 hr at a power level of 10 Mw). The specimens have been disassembled, but testing will not begin until after examination with an optical comparator to separate the bent cnes. Limited creep and tensile tests will be run to determine whether the properties fall in line with those determined previously for material irradiasted to the same dose. The second objective of this program has been to understand the mechanism of the damage in Hastelloy N and to determine a means of making a basic Improvement in its resistance to irradiation damage. We have found that the damage is dependent on the thermal-neutron flux. This is illustrated by the data in Table 6.2. Specimens were exposed to neutron enviromments of various energles, and it was found that the damage was fairly independent of fast dose and depended primarily on the thermal dose. It is believed that the specific thermal-neutron re- action is that of the transmutation of °B to helium. 119 Table 6.2. Tensile Ductility of Hastelloy N (Heat 5065) at 760°C & = 0,002 min™*t o, =1 X 108 nvt, 0., =1 x10%% ave, o =5 X 10 nvt, éf = 5 % 10 nvt ¢f =1 X 10*8 nvt ¢f =1 X 10 nvt 7.1 6.2 12.0 Hence, the first approach Lo the problem has been that of reducing the boron level. Figure 6.19 shows the postirradiation creep-rupture properties of several air-melted heats of Hastelloy N. 'The materials contain between 20 and 40 ppm boron. The rupture ductilities are given in parentheses. There seems to be little effect of irradiation temper- alure on the rupbure life or ductility. Figure ©.20 shows the proper- ties of two vacuum-melted heats containing 7 ppm (2477) and 9 ppi horon. (65-552). When the materials are irradiasted at a low temperature, theilr properties excel those of the alr-melted heats. When these same heats are lrradiasted hot, they exhibil properties comparabvle with those of the alr-melted heats. ORNL-DWG 86-11453 70 - T T T T T — - L | ] N 1T Tfl 80— \\%Sjnidmr(bu)“ T N 50_ 1 1 1 -1 ‘ - TYPICAL C ELTED — 40 — Fwmwos - o e N 2 [ ‘ e U’) v W . ! E i ‘ (2.0 = i » 30 |— uig | - ALLCYING HEAT NO. ELEMENT o 21541 W ® 24548 - ANNEALED 100 hr 20 1421542 W AND Nb | AT 874°C PRIOR TO [ = S e e e A 21545 Ti IRRADIATION 021543 Nb ®65-552 — WORKED AT 871°C o || IRRADIATED AT €50°C B o <1>m=3.5x1020 vt OL _____ | m I _,L\ i m \J \ L o 1 10 100 10,000 RUPTURE LIFE (hr) Fig. 6.19. Postirradiation Creep of Several Alloys at 650°C. Num- bers indicate fracture elongations. 120 ORNL-DWG 66-10882 o T T T T 1T -]CLOSED POINTS: COLD (RRADIATION OPEN POINTS: HOT IRRADIATION | 1] 1 L 60 lm 2 —— - i ! | ] //,UNRRADMTED PROPERTIES i | T < J\” _ 40_#‘ | W N J‘ ; | : 2.2 |31 _L Q ! | . o | ¢ s ™ T 30 - e o 34 3.2 v f { ! il | 20 —— DOSE IRRADIATION e ! HEAT NO. ®,,= TEMPERATURE (°C) J . ’- o 2477 5x 1020 43}‘ | ‘ 10l ¢ 65-552 5x1020 150 i | ] A 2477 251020 650 | | O 65-552 2.6x10°° 650 | | ANNEALED 1hr AT 4477 °C PRIOR TO | | ¢ 2477 5 %1020 150 | | IRRADIATION |‘ ! | ! 0 e vl | LU 1 10 100 1000 10,000 RUPTURE LIFE (hr) Fig. 6.20. Postirradiation Creep-Rupture Properties of Several Vacuum-Melted Heats of Hastelloy N at 650°C. elongations. Numbers indicate fracture Several of the alloys were irradiated Lo various doses, and the helium content was calculated from the dose and the original boron con- tent of the alloy. Figure 6.2l shows the rupture ductility in a tensile test at 650°C and 0.002 in./min as a function of helium content. There is considerable deviation about the line drawn, with some of it probably being significant. The two vacuwm-melted heats (65-552 and 2477) again show lower ductilities when irradiated hot. The two Llow points for heat 4065 indicate a possible influence of high irradiation temperature on this heat. However, the most important observation is that, for a rea- sonsble *°B content and a reasonable lifetime, the helium content will be between 1072 and 107% atom fraction. The variation in ductility over this range of helium contents 1is only aboub 25%, agsuming that the line drawn 1s reasonably correct. A similar plot is shown in Fig. 6.22 which relates the helium pro- duced in the alloy to the fracture dvctility. The line drawn seems 4o be representative of the higher-boron air-melted heats (5065, ete.). The vacwm-melted low-boron heats (65-552 and 2477) exhibit better duc- tility if they are irradiated cold but actually have lower ductility when irradiated hot. Hence, the role of boron in these alloys is not at all clear., We have not prepared a serles of alloys where everything elge has been held constant except the boron content. 1L seems that the variations 1in composition and melting practice must influence the ORNL-DWG 66-11454 LT ETR 41-30 BSR ETR 41-3f ORR -149 ORR -155 ORR-153 ANNEALED 1 hr AT 1177 °C PRIOR TO IRRADIATION ’ TESTED AT 650 °C, 0.002 min~" : | ~ GLOSED POINTS: COLD IRRADIATION “ \ 1 i | \ | i TOTAL ELONGATION (%) & OPEN POINTS: HOT IRRADIATICN el 1078 ol 107 ? 1077 He CONTENT (otom fraction)} Fig. 6.21. Variation of the Postirradiation Tensile Properties of Jeveral Heabs of Hastelloy N with Helium Content. ORNL-DWG &6-10880 Fig 14 I \ T T T { CLOSED POINTS: COLD IRRADIATION | A ORR-153 | OPEN POINTS: HOT IRRADIATION 0 BSR [ ey L B : i 2477 ORR-141 UNIRRADIATED v2d . : ORR-148 DUCTILITY N M | ANNEALED 1 hr AT 10 poormrmeto R 4 1H 4177 °C PRIOR TO- l } IRRADIATION _ \ ~\ L || TESTED AT 650 °C Hi > 8 \ kwl WI” .- L 1‘ ‘ ‘_&* S 5 T 2 2477 Ny 5065 C ol AEPY ‘ 324 | L ) | | 5065 . alo— L. i 400 \hl\ \ ‘ ’ ‘ 65-552 \lg | 3247 it LL : || 2 | ‘ | | 324 \ : i ‘ ‘: { 4q -.4 \ o - l 1078 1" - He CONTENT ({atom fraction) O.22. Variation of Postirradiation Creep Ductility of Hastelloy N with Helium Content. Numbers indicate stress in ksi. 122 distribution of boron or the effect that the transmuted helium has on the properties so greatly that the role of boron per se is masked. The great effect of irradiation temperature on the postirradiation proper- ties of the vacuum-melted alloys supports this supposition. Although we have known for some time that the ductility of an ir- radiated test specimen varies greatly with temperature, we have made extensive use of postirradiation tensile tests for screening purposes even when we were interested in long-term applications. Figure 6.23 showeg how misleading this approach can be. The rupbture ductility in postirradiation tensile and creep tegtse is plotted to illustrate the influence of zirconium content on the properties. The two sets of points at the left were obtained by tensile tests at two different strain rates (0.05 and 0,002 in./min). The trend is evident that the ductility improves with increasing zirconium content. However, the duetility in the creep tests (represented.by*the polints at the right) shows no systematic dependence on the zirconium content. This material contained extremely high boron (approximately 200 ppm), and we believe that the data do not warrant the conclusion that zirconium has no beneficial influence. The important point is that tensile tests may not be adequate even for screening purposes in nickel-base alloys. The more expensive creep test may be necessary. ORNL—DWG &65--10885 T I | T TETTT [ | E? Zriwt %) ® — ° 0.05 T . 013 { [l:2 18 - Lgd | 1| & 0.43 | REMELTS OF HEAT .| g o 0.52 NO. 5065 ’ * " 08 ‘ 1 v 44 12 | — . Ty 1.2 @, =2x1020 A/ . IRRAD TEMPERATURE = 650°C 1 L T P TEST TEMF’ERATUREZGSO"C - . ELONGATION (%) @ RUPTURE LIFE (hr) Fig. 6.23. Influence of Zirconium Content on the Postirradiation Ductility of Hastelloy N. 123 Several small commercial melts have been procured and evaluated. The basic alloy is Ni—12 Mo~7 Cr—0.05 C, the reduction in molybdenum content being made to suppress second-phase formation. The alloy addi- tions investigated include nicbium, tungsten, and titanium. To date, these alloys have been investigated in only one metallurgical state, 40% cold working followed by 100 hr at 871°C. This treatment produces a very fine grain size and probably does not result in the optimum properties. Figure 6.24 shows the postirradiation creep properties of several of these alloys at 650°C. The very encouraging alloy is No. 21545, which contains 0.5 wt % titanium. The rupture life is better than the other alloys, but the rupture ductility is far superior. Figure 6.25 shows a compilation of all the in-reactor creep tests run on various heats. All the data are contained in a scatter band bounded by heats 5065 and 5085. With few exceptions, all the heats ex- hibit fracture ductilities of 1 to 3%, with no systematic heat-to-heat variation evident. Three of the heats lend support to the idea that there may be some critical stress below which irradiation damage be- comes minimal. One must face the question of why heat 21545 locks ex- cepbionally good in postirradiation creep tests (Fig. 6.24) and not in in-reactor creep bests (Fig. 6.25). With a closer examination, one sees that the test at 21,000 psi (Fig. 6.25) did not fail during the experi- ment. This point indicates that the threshold stress for damage 1n this alloy may be higher than that for the other alloys (approximately 15,000 psi)., Thus the in-reactor creep tests do suggest that the titanium- bearing alloy may be superior. ORNL-DWG 66-10879 1 ‘ ‘ ° N | T T T T T T 1T e ‘ ! ALLOYING AN HEAT NO. ELEMENT 14 |- : o 21541 w 4 N a 21546 — . o 21542 W AND Nb g;l;{g%g%éogom Al \ ¢ 21545 Ti {RRADIATION 12 T N T A 21543 Nb \ \\‘ \, a 65-552 - WORKED AT 871°C N N 10 7\ N .. IRRADIATED AT 650°C i E ;§\ \ N Gy = 3.5 X AT vt z \ % ™ l S 4 N N = NG T —t i < N A 2 N ™ fi\\ 21545 O i N | ‘ d 6 ___B.u. \ \3‘ “ 1. \"""-- ] \\ N \4\\~ N ‘ N \ TH\\ . | 4 - \\ \J\ N e, \‘ ______ \fi \ \\\\\ / ~ % e ™ | \ o / 65-552 ST X ! 1 \ A | - Nl LT g o AL o LI oo 04 4 10 100 RUPTURE TIME thr} Fig. 6.24. Postirradiation Ductility of Several Alloys at ©50°C. 124 ORNL-DWG 66-11473 70 . - - TNea T T | N | \\ ' ' 60 1 | E S - e TYPICAL EX-PILE P RT \\\\ /////,, ‘ 1 }ROPER IES \.<"“ 50 bomrm e L . 1] \\\ & Qo ! o 40 -~ - o e = ! @ \\ta.zl:: \ 30 (13)+,N- ---------- ‘ U'_;. (10 ‘ \ ‘ | - ALLOYING | H Hotedsin .9 \\ (125 201 HEAT NO. ELEMENT | 1 e {22 gzfinm—mqm ot o 2065 _— fhr AT 1177 °C LR N ® 5085 e ‘ (1‘0‘)‘ (2.8)N(4.6) e \\ o 7304 e fhr AT 4177 °C+8hr AT 871 °C o s A s 51245 VAP VL roonat 871 oc | o | 10 |- & 21545 ! | ‘ l ‘ IRRADIATED AT 850 °C r ‘ Kfim= 6x10"3 v | RUPTURE DUCTILITY IN ( ) L J L I | . | 1 10 100 1000 10000 RUPTURE LIFE (hr) Fig. 6.25. Comparison of In- and Ex-Pile Creep Rupture Properties of Hastelloy N at 650°C. ORNL-DWG 66-10878 1800 (- *T—— ] | ' e [ e i i 1600 | I ‘ s 53 14900 1 — Ll L #100 l 58 ’ ’ 1200 |— | | £ , ‘ W 1000 L | ANNEALED 1 hr AT 4{77°C L PRIOR TO IRRADIATION Lyt % 800 [#12.5 —— —— IRRADIATION TEMPERATURE=650°C- £ UNIRRADIATED, Dy = 2.5 X 1070 s 5 STD HASTELLOY N ™ 600 - o572 | TEST TEMPERATURE=650°C T STRESS=32,350 psi 400 { | J 200 | g o —— P o4 ‘ ’ O ‘ | ,,J ........... L ,,,v....L,,,,,,,..._L..._______...‘.,,i O ©Ot 02 03 04 05 06 07 08 09 10 Ti CONTENT (wi %) Fige. ©.26. Influence of Titanium on the Postirradiation Creep Prop- erties of Ni—12 Mo—7 Cr-0.05 C. Numbers indicate fracture elongations. 125 Figure 6.26 shows the results of postirradiation creep tests on sev- eral laboratory heats containing various amounts of titanium., All the alloys are superior to irradiated standard Hastelloy N with respect to both rupture life and ductility. Several of the alloys exhibit rupture lives in excess of that of uwnirradiated Hastelloy N and ductilities as high as 10%. Characherization of Hastelloy N for Service at 982°C - H. E. McCoy Since Hastelloy N is being consldered as the structural material for a molten-salt distillation vessel, the Mechanical Properties Group wag asked to determine the creep properties of this alloy at 982°C. The results of this brief study are presented, and several potential problems associated with the use of Hastelloy N ab such an elevated temperature are polnted oub. Although the vessel presently being built will be used only with nonactive salts, it is anticipated that a similar distillation apparatus will be an integral part of a Molten-Salt Breeder Reactor. For this reason the questions concerning the use of Hastelloy N for this application should ve resolved., The creep~rupbure properties of Hastelloy N are well documented® 7 over the temperature range of 600 to 800°C where it is commonly used. However, data at 982°C were nonexistent, since this is above the normal service temperature for this alloy. The results of a brief creep-rupture program undertaken to supply data at 982°C are shown in Table 6.3, Plots of these game data are shown in Figs. 6.27 to 6.29. These data were ob- tained on test specimens of a typical heat of air-melted Hastelloy N (heat 5065). The test specimens were small rods having a gage section 1 in. long X 0.125 in. in diameter. All tests were carried out in dead-weight creep machines in air. The creep properties shown in Fig. 6.27 indicate that the stresses to produce rupture and 5% strain in 1000 hr are rea- sonably well defined. However, the stresses for 1 and 2% strain in 1000 hr are not defined sufficiently by experimental data. The minimum creep rate data in Fig. 6.28 correlate very well, but these numbers are not very useful for design purposes, since the minimum creep rate only applies during a small fraction of the life at a gilven stress. The plot of the Aductility shown in Fig. €.29 exhibits a ductility minimum for test spec- imens having rupture lives over the approximate range of 50 to 200 hr. However, the minimum ductility observed still is slightly in excess of 20%. The reduction in area decreases rapldly with increasing rupture life to a value of about 15%. The increase in the elongation for long rupture lives is thought to be associated with the very extensive in- tergranular cracking that cccurs. The specimen from test 5768 was examined metallographically. The extent of intergranular cracking is illustrated by Fig. 6.30. It was also noted that the quantity of precipitate present after testing was much greater than that present before testing. Filgure 6.31 shows a typical area of this heat of Hastelloy N prior to testing. Tigure 6.32 shows the tested specimen. The precipitate marked "A" is believed to be that present in the starting material. The larger precipitate, marked "B, " is thought to be induced by the thermal and mechanical history. Table 6.3, Creep Properties of Hastelloy N at agz°c® Time to Indicated Strain (hr) o G Che ) Teme : 1% 2% 5% 20% Rupture 5768 2,000 35 73 185 495 649 0.0276 41.50 16.22 5765° 3,000 1 7.6 33 117 123.9 0.123 22.36 13.4 5762° 4,000 0.7 2.0 11 47 51.5 0.282 21.88 15.47 5761 6,000 0.75 1.3 3.2 10.5 16.15 1.65 48 . 4ds 36.57 5763 10,000 <0.1 0.2 0.5 1.9 2.95 9.0 50.0 37.86 5867 1,880 35 72 165 407 520.8 0.026 52.0 15.6 5868 3,000 12 25 60 154, 157.9 0.081 22.22 15.6 91 “Heat 5065 tested in the as-received condition (1/2 hr at 1176°C mill annesl). Temperature control not adeguate at start of test. 127 ORNL-DWG 66- 11455 20,00G |-y - ]H\ HASTELLOY N HEAT 5063 10,000 5000 Fig. 6.27. Creep-Rupture Prop- ------ T erties of Hastelloy N at 982°C. STRESS ( psi ) 2000 ; o RUPTURE 1000 O.1 i 10 100 1000 TIME {hr) ORNL-DWG 6611458 10,000 " HASTELLOY N HEAT 5065 - 982°¢C 5000 |- L L. e . o . a Fig. 6.28. Minimum Creep Rate = w vs Stress for Hastelloy N at 982°C., W > 2000 1000 0.0 0.1 { 10 MIN CREEP RATE (% /hr) ORNL DWG 686-11457 6O [ ; : 60 | | | | 3 50 - : : / 50 240 |— b ‘ ‘L e /o a0 — T = 2 | 2 an j | 3O< Fig. 6.29, Ductility of + T EE - et = . | "\ = Hastelloy N at 282°C. m . > HEAT 5065 \ ‘ e a O ELONGATION bl L[| 2o 2 Z 20 - ] 20 2 A REDUCTION IN AREA 2 DUCTILITY OF HASTELLQOY ~ Al N AT 982°C a 10 F—1 g | 10 1 100 1000 RUPTURE TIME (hr) 128 .U INCHES M 10CX Fig. 6.30. Photomicrograph of Hastelloy N Tested at 982°C and 2000 psi. 0.035 INCHES 10CX M s Fig. 6.31. Photomicrograph of Hastelloy N, Heat 5065, in the As- Received Condition. 129 - Y-74240 0.035 INCHES M 1oCxX [_.. iCn £ 500X 0.007 INCHES Fig. 6.32. Photomicrographs of Hastelloy N Tested at 982°C and 2000 psi. 130 Staining with a KMnO4-NaOH solution indicated that the precipitates were of two different compositions. Although this precipitation dces not ap- pear to impair the ductility at 982°C, it remains to be established whether the low-temperature ductility is affected. BSince the vesgsel will be ex- posed to numerous thermal cycles, the high- and low-temperature ductilities are both of interest. The results of thisg study are hardly adequate for the design of a distillation system that is to be an integral part of a reactor system. This study has shown the need for (1) strength data extending to longer times, (2) tests to evaluate the influence of the precipitation on the ductility, and (3) oxidation data under conditions of constant and cyclic temperatures. 6.8 Thermal Convection Loops A, P. Litman G. M, Tolson We are continuing to study the compatibility of structural materials with fuels and coolants of interest to the Molten-Salt Reactor FProgram, Natural-circulation loops described previouslyl8'19 are used as the standard test 1n these studies. Currently, four loops are in operation. Three thermal-convection cir- cults containing simulated MSRE fuel salt and fabricated from either Has- telloy N, type 304 stainless steel, or Wb—1% Zr alloy with type 446 stain- less steel external cladding have operated for approximately 4.5, 3.2, and 0.6 years respectively. Operating conditions for the loops are de- tailed in Table 6.4. The Hastelloy N and type 304 stainless steel loops have continued to circulate salt without incident. However, the refrac- tory-alloy circuit has shown some degradation of late. This was evi- denced by the cold leg gradually losing temperature despite the addition of insulation. It is suspected that the smaller internal diameter of this loop (0.3 in.) has contributed to the operational problems. The fourth operating loop (loop 10) is fabricated from lastelloy N and contains a proposed secondary coolant (NaF-KF-BF3, 48-3-40 mole %) for the reference-design MSBR. This circuit has operated for over 3000 hr with the hot leg at 1125°F and a temperature differential of 265°F. During this reporting period, a Croloy M loop (loop 12) circulated the proposed secondary MSBR coolant for 1440 hr, after which time it was shut down due to plugging. X rays taken of the loop disclosed spotty high-density regions in the cold leg. A few suspicious areas were also seen in the hot leg. The loop piping and heat-transfer fluid will be sub jected to complete chemical and metallurgical analysis. A Croloy 9M (loop 8) natural-circulation loop containing lead with 230 ppm magnesium as an inhibitor plugged after operation for 2950 hr. A section through the cold leg in the plugged region is shown in Fig. 6.33. Examination of the loop components is proceeding. Table 6.4. Thermal Convection Loop Operation Through September 30, 1966 Hot-leg Maxdimum AT Time Loop Material 5 . Heat Transfer Medium Temp. o Operated pecimens o % (°F) - (°F (tr) Hastelloy N Hastelloy N + LiF-BeF 5 ~Zrf ,-UF,-ThF, 1.300 160 39,400 2% Mo (70-23-5-1-1 mole %) Type 304 stainless steel None LiF-BeF »-ZrF ,-UF 4~ThF ., 1250 180 28,125 (70-23-5-1-1 mole %) Type 446 stainless-steel- None LiF-Belp-ZrF 4~UF, 1400 300 5,235 clad No=1% Zr (65-29.1-5-0.9 mole %) Hastelloy N None NaF -KI'-EF 5 1125 265 3,110 (48-3-42 mole %) Croloy 9M Croloy OM Nal -KF-BF 5 1125 260 1,440 (48-3-49 mole %) Croloy 9M Croloy OM Po + 230 ppm Mg 1100 200 2,950° %Ioop plugged on 9/26/66. Loop plugged on 6/9/66. TeT 132 PHOTO 73481 Fig. 6.33. Matching Halves of Portion of Cold ILeg from a Croloy OM Loop Which Circulated Lead with 230 ppm Magnesium as Inhibitor for 2950 nr at 1100°F, To date, all uninhibited lead systems constructed of carbon, low- alloy, and stainless steels have tended to plug due to the formation of dendritic crystals of iron and chromium in the cold regions of the loops. The hot-leg attack has consisted of uniform surface removal with isclated pits extending to a greater depth. While Nb—i% Zr alloy has exhibited no measurable hot-leg corrosion during test, niobium crystals have been found in the cold leg of a loop which operated for 5000 hr at 1400°F. All these findings are discussed in detail in a report sumarizing re- sults to date on the circulating-lead loop program.ao References 1. MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, pp. 87—92. 2. MSR Program Semiann. Progr. Rept. Feb., 28, 1966, ORNL-3936, pp. 107-8. 3. W. P. Etherly et al., Proc., U.N. Intern. Conf. Peaceful Uses At. En- ergy, 2nd, Geneva, 1958, 7, 38901, 4. W. Watt, R. L. Bickerman, and L. W. Graham, Engineering 189, 110-11 (Janvary 1960). 5. C. R. Kennedy, Metals and Ceramics Div, Ann. Progr. Rept. June 30, 1965, ORNL-3870, pp. 194—97. 10. 11. 12. 13. 14. 15, 16. 17. 18. 19. 20, 133 R. G. Anderson and J. ¥. W, Bishop, "The Effect of Neutron Irradia- tion and Thermal Cycling on Permanent Deformations in Uranium Under Load,"” pp. 17-23 in Uranium and Graphite, Monograph 27, The Insti- tute of Metals, London, 1962. J. C. Bokros and R. J. Price, Radiation-Induced Dimensicnal Changes in Pyrolytic Carbons Deposited in a Fluidized Bed, GA-6736 (November 1965). To be published in the Journal of Applied Physics. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, p. 104. R. G. Donnelly and G. M. Slaughter, Welding J. fi;fl5), 461-69 (1962). MSR Program Semlann. Progr. Rept, Feb. 28, 1966, ORNL-3936, pp. l O]...-‘.Zl' . W, R. Martin and J. R. Welr, Effect of Elevated Temperature Irradia- tion on the Strength and Ductility of the Nickel-Base Alloy, Hastel- loy I, ORWL-TM-1005 (February 1965]. W. R. Martin and J. R. Weir, Postirradiatlon Creep and Stress Rupture Properties of Hastelloy N, ORNL-TM-1515 (June 1966). MSR Program Semiann. Progr. Rept. Aug, 31, 1965, ORNL-3872, pp. 94— 1.05 * MSR Preogram Semlann. Progr. Rept. Feb. 28, 1966, ORNL-3936, pp. 111~ 21. D. R. Harries, J. Brit. Nucl. Energy Soc. 5(1), 74 (January 1966). Jo. Te Venard, Tensile and Creep Properties of INOR-8 for the Molten- Salt Reactor Experiment, ORNL-TM~1017 (February 1965). R. W. Swindeman, The Mechanical Properties of INOR-8, ORNL-2780 (Jan. 10, 1961). G. M. Adamson, Jr., et al., Interim Report on Corrosion by Zirconium Base Fluorides, ORNL-2338 (Jam. 3, 1961). MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, pp. 81— g7. G. M, Tolson and A. Taboada, A Study of the Lead and Lead-Salt Cor- rosion in Thermal-Convection Loops, ORNL-TM-1437 (April 1966). 7. CHEMISTIRY 7.1 Chemistry of {he MSRE Behavior of Fuel and Coolant Salt — R. E. Thoma Several refinements of analytical methods were made during the low- power operating period of the MSRE which led us to believe that the composition and purity of the reactor salts can be ascertalned accurately and economically on a routine basis.’ Within the last few months we have - attempted to confirm this belief by demonstrating that samples of the reactor salts could be obtained and analyzed on a routine basis with the reactor operating at full power. In addition we have sought to establish from the results a practical optimal frequency of sampling. Comparisons of the current data with previous results indicate that the fuel and coclant salts have not undergone perceptible composition change since they were first circulated in the reactor some 16 months ago. The con- centration of corrosion products has not increased appreciably in these salts within the report period. fuel Salt Composition and Purity. Twenty-~five samples of fuel salt were removed from the MSRE during full-power runs FP-5, -6, and -7 for compcosition anaLysis.z The results of thnese analyses, listed in Tables 7.1 and 7.2, show clearly the chemical stability of the fuel solution; it has maintained constant composition since it was constituted in the reactor in 1965. Structural-Metal Impurities. The concentrations of structural-metal corrosion products found in the MSRE fuel salt in the period January to August 1966 are compared with previous assays in Table 7.3. Since chro- mium is the most chemically active constituent of Hastelloy, all cor- rosion reactions should, on proceeding to equilibrium, result in the . production of CrFo. The extent of corrosion in the MSRE is currently monitored by analyzing for the chromium content of the salts. In its full-power operation during the period PFebruary to July 1966, the MSRE has behaved well with respect to corrosion chemistry; no apparent cor- rosion has occurred within the fuel or coolant systems. We have con- cluded in the past that corrosion might be influenced by the presence of a small amount (perhaps approximately 1%) of UF; formed in the IilF- UF, enriching salt during preparation. When present, UF; is readily oxidized to UF, in reaction with CrFj: CrlF, + UF3 = cr® + Uk, . The presence of MoF, or Mol's in the fuel salt, as inferred from the re- sults of recent examinations of surveillance specimens removed from the reactor core,3 is puzzling in connection with the likelihood that UFj sti1ll remains in the fuel salt. The notable freedom from corrosion during this period therefore is reassuring. 135 Table 7.1, Results of MSRE Fuel 8alt Analyses, Runs FP-5, -6, and -7 Concentration (wt %) u;gple Mirhr - . Li Be Zr U* F 5 FP5-1 26.5 10.65 6.53 11.45 4625 68,18 101..44 FP6-2 11L.43 D.22 10.80 4.622 68.97 102,04 TP6=5 10,45 6.37 11,12 4,605 69,85 102.40 FPe-6 10.45 646 11.32 4625 68.56 101.52 Pe-7 180 10.32 6 41 11.42 4o 047 69 .67 102.47 FPo-8 10,43 649 11.2%9 4655 68,20 101.77 FP6-9 10,48 6.066 11.52 4 684 68.92 102.26 FP6-10 10.30 6. 70 11.54 4,595 685,81 102,00 FPe-11 10.30 6442 11.28 4,012 70,02 102.70 Fre-13 1000 10.50 6.41 11,06 4.628 66.79 99,38 Pe~14 10.50 645 11.33 4,617 68,18 101.08 TPG6-15 10.55 6.86 11.35 4 601 68.1.8 101.54 P6-16 10.55 6.58 11.31 %« 629 67.88 100.90 FPe-17 11 .44 O .64 11.05 4.652 68.26 102.04 FP6-19 10.40 6.88 11.72 4667 69.1.0 102.77 FP7-1 2920 10.60 6.78 11.16 e g GA4TT 69.26 102.45 FP7-3 10.55 6.56 11 .44 4 .656 68 .24 101 .45 FP7 -4 10.50 6.40 11.61 4 640 67.32 100,97 FP7-6 4626 10.60 6.63 11.35 4614, 69.22 102.41 Fp7-7 10.55 6.65 11.13 4 o 641 69.60 102,57 FP7-8 10.60 .59 11.67 4,663 67.87 101.39 FPr7-10 10.63 6.91 11.21 4 . 609 68.20 101..56 FP7-11 6900 10.45 7.00 11.22 4 .630 69.48 102.78 FP/7-12 1.0.50 6.65 11.04 4 o 640 68 .44 101.27 FP7~14 10.55 6.50 11.26 4 .660 68.79 101.76 ¥FP7-1.6 7823 10.55 6.71 11.60 4,638 67.59 101.08 The average concentrations of the structural-metal impurities which have been found in the MSRE fuel sall since the beginning of reactor op- erations are given in Table 7.3. An interesting trend is evident in the above results. As impurities, iron and nickel are presumed toc exist as colloidally dispersed metallic particles. The concentration of iron has decreased slowly bubt steadily during reactor operation, while nickel analyses remained essentially constant. Oxide Analyses. FEight samples of MSRE fuel were analyzed for oxide content during the recent full-power operations. etails of the proce- dures employed are given elsewhere.® The average value of oxide concen-~ tration was found to be 54 ppm, significantly lower than that found in 136 Table 7.2. Summary of MSRE Fuel Composition Analyses Component. Nominal Boock FP4 P5-7 FP4~7 Mole Percent LiF 65,00 64 . 88 63.36 £ 0,567 62.20 £ 0,764 ©3,29 * 0,721 BelF o 29.17 29.26 30.65 + 0.583 30.76 = 0.746 30.70 + 0,695 Zriy 5.00 5 .0¢ 5.15 = 0,116 5.22 £ 0.129 5.19 £ 0,126 UF 4 0.83 0.82% 0.825 + 0,011 0.822 + 0.011° 0.824 + 0.0LL Weight Percent Li 10.95 10.93 10.51 + 0.137 10.54 * 0,017 10.52 +* 0,178 Be 6.32 ©.34% 6.55 + 0,161 6.60 * 0.183 6.57 * 0,179 Zr 10.97 11.06 11.14 £ 0,295 11.33 % 0.220 11.24 £ 0,271 U 4,73 L6462 4642 t 0.028 4.635 t 0.022° 4.638 0.025 9237.003 = 33.241 wt % 235U, ©237.009 = 33.06 wt % 235U. Table 7.3. Concentration of Cr, Fe, and Ni in MSRE Fuel Salt Run Nugber of Concentration (ppm) No. Samples O Fe Ni. ¥p-3 51 37 £ 8 154 + 55 48 * 19 Fr-4 22 48 + 7 131 + 65 20 £ 20 rp-5,6 14 50 £ 7 108 * 44 54 * 25 rp-7 11 48 + 6 79 + 38 48 + 23 the zerc-power experiments (95 ppm). Since fluorine can be evolved from frozen irradiated fuel, it was necessary to consider the possibility that oxlde is released by the reaction 7S + 0,7 = =08 + 2F . o Sample specimens were analyzed in one instance aflter minimum delay {ap- proximately a 7-hr interval between sample removal to HF purge) and in another instance after a 48-hr delay. Results of the two analyses were coincident within experimental error. Although the minimum delay period is probably not sufficilently long to allow any of the oxide to he con- verted to oxygen, plans are under way to test the infrinsic capability of the method by standard additions of oxides to future samples.4 137 Table 7.4. Isotopic Composition of Uranium in MSRE Fuel Salt Specimens Isotopic Composition of U (wt %) Sample Mirhr No. 2347 235y 236(; 238y #P2-10-13 0 0.356 33.716 0.145 65,783 Fre-14 1000 0.352 33.534 0.144 65.970 FP6-19 2900 0.345 33 Ak 0.151 66.060 FP7-8 5000 0.352 33.400 0.163 66.085 FP7-12 7000 0,349 33.329 0.174 66.148 FP7-15 7600 0,347 33,161 0,177 66,315 Isotopic Analysis of Uranium. Varistions in the relative concen- trations of uranium isotopes in the MSRE fuel are determined on a regular basis by mass spectrometric analysis.5 A summary of the isotopic analyses of the MSRE fuel is given in Table 7.4, which shows qualitative evidence of 23y burnup and ingrowth of 238, The limit of experimental accuracy (#0.1% of individual values) and the low burnup fraction (approximately 1%) render the current values of little use for accurate caleculation of M5RE power levels at present. 1Tt is anticipated that as a greater fraction of *3%U is consumed, such determinations will assume greater significance in corroborating burnup computations which are based on rerformance data. Coclant Salt Composition. Coolant salt was circulated in the MSRE for some 1200 hr during the prenuclear test period. Since flush and coolant salts were supplied from the same reservolr, the coolant salt was analyzed during this period only for impurities. At the end of pre~ nuclear testing, concentrations of the structural-metal impurities, Cr, Fe, and Ni, were established to be approximately 30, 90, and 7 ppm, respectively, which are remarkably low considering that the circuit had not been Tlushed previously with molten salt. The results of spectro- chemical analyses made at that time showed that no significant amounts of additicnal impurities were introduced during this period.6 As the reactor was brought to [ull power early in 1966, coolant salt was again circulated, sampled regularly, and subjected to compositional and im- purity analysis. The results obtained from these analyses are given in Table 7.5. In addition to the components, ceolant salt was analyzed for, and found free of, zirconium by wet chemical and spectrochemical methods. The intention was to confirm that no fuel salt had been transferred to the coolant circuit. Analyses of coolant salts are conducted in the General Analysis laboratory of the ORNL Analytical Chemistry Division, while those for the fuel are obtained from the General Hot Analyses laboratory of that 138 Table 7.5. Results of MSRE Coolant Salt Avalyses, Runs CP-4 to -7 Concenbration 1 Total Time S;mp © in Pump Bowl Wt % ppm e (hr) Li Be F 5 Cr Fe i o CP4-1 36 65 <5 CP&4-2 11 26 83 <5 130 CP4~3 59 13.78 .91 76,70 99.39 41 41 37 185 CPL—d 67 53 46 24 CP4-5 180 14.20 g8.87 '76.80 99.87 50 50 20 CP4~6 229 38 CP5-1 13.86 8.85 76,7 29.41 35 5 <10 150 CP5-2 13.82 8.55 76.6 98.97 35 <2 <10 110 CP5-3 12.76 10.51 76.04 99,31 46 57 64 65 CP5-4 13.00 9.17 76.8 98.97 35 119 14 <20 CP5~5 12.60 9.15 76.7 98.45 42 87 28 1.20 CP5-6 13,01 10.06 77.2 100,27 20 100 8 260 CP5~7 855 13.17 9.49 76.4 99.06 24 112 <2 171 CP5~8 12.70 10,22 772 100,12 29 36 0 4 = az . ] Lo 5 . 1 PuvP oF o ] X0, Q << I Bi nqz l__ ) =0 S the extraction vessel. Thus, at low flow rates thorium would be introduced at a rate regulated by its solubility in bismuth. The alloy was then sprayed into the salt mixture so the protactinium would be extracted at the surfaces of free-falling droplets of bismuth. A pseudo-first-order rate of extraction was eXpected. Although the blanket reprocessing method involves stripping the prot- ) actiniun from molten bismuth into a second salt mixture by hydrofluori- ) nation, this feature was not included in the initial pump-loop design. Instead, the recirculating molten-metal stream was pumped through a bed - of steel wool to provide for the collection of protactinium by absorption on the iron surfaces or by filtration of suspended particles. This ex- pedient was based on results of the experiments described above and on another in which thorium metal was added to a blanket-salt mixture that contained ???Pa. The protactinium was found uniformly distributed on steel wool that had been immersed in the salt. Accordingly, the steel wool column was designed to provide a largs surface area of 1lron relative to iron surfaces exposed 1o the liguid phaseg elsewhere in the system without unduly restricting the flow of bismuth. The extraction vessel was also fitted with an open cylinder of niocbium for primary coatainment of the salt mixture. The centrifugal pump was the same as that designed and operated by F. S. Bettis, Reactor Division, 1n similar molten-salt—molten-lead systems. Pump-Loop Operation. Several related experiments were carried out in the pump loop. The loop was first charged with 13.5 kg of bpismuth that 151 had been previously treated with hydrogen at 600°C for oxide removal. This material was circulated through the system to ascertain operational pro- cedures and pump performance characteristics. The steel wool colunr was then prepared on its fixture, inserted into the system, and fired with hydrogen at 700°C in situ while bismuth was static, and the balance of the system was protected from oxide contamination by flowing helium. The column contained 12.5 g of grade 1 steel wool having a total surface area of 0.49 m? or about ten times that of the geometric surface of iron that was elsewhere exposed to bismuth in the circulating system. The salt mixture IiF-BeF,-ThF, (73-2-25 mole %) was spiked with about 1 me of 233pa from irradiated ThO, by our usual HF-Hp treatment in nickel at 600°C and with Hs; alone at 700°C. Approximately 6.7 kg of this material was transferred into the niobium-lined extraction vessel. Bismuth was at first circulated without a reducing agent, to estab- 1ish the stability of 233Ps in the salt mixture. Then thorium metal was introduced in small amounts by submerging a basket of thorium chips in the bismuth stream at the outlet of the column. Five successive additions of thorium were made in this manner during this phase of the experiment. The radiochemical results obtained from filtered samples of the two liquid phases during some 30 hr of pump operation are shown in Fig. 7.8. It is interesting to note that the 233pg, activity in the sall phase was exX- tremely stable during the first 17 hr of loop operation and the first three additions of thorium. After the fourth addition of thorium, the 233py activity in the salt phase decreased at a measurable rate until apparent exhaustion of the thorium metal and continued after the fifth thorium addition until 57% of the *°2Pa was removed. As shown by Fig. ORNL-DWG 656-11465 23300 ACTIVITY FOUND (%) 120 e £ - o 100 ¢ - 5 e e o » 2 © SALT PHASE Z 80 % o METAL PHASE e ) i 233 . 5 Fig. 7.8. Behavior of Pa in ~| Pump Loop During First Extraction Ex- w - . 6o | I B . 5l periment (Th® Added to Bismuth As ! =z ? ? \ 0| Noted). . = = z | . 5 g8 o 3 z £ o < <1 < . a _ 2 o M O_c < o.c HDI: g‘; < 8 Ew Fe FZrx o8 . - o £ =0 o b & b = g 20 e ' — s F] i ) | ' ¢ # + - T L ® : = os ; J ./ Ok L. » o L e—fpe-s-a— ¢ —o @5 * O oo Q 5 10 15 20 25 30 LOOP OPERATING TiME (hr} CRNL-DWG 661466 Wi | Fig. 7.9. Loss of 22%Pa from LiF-BeF,-ThF, (73-2-25 Mole %) by Reduction with Thorium in Bismuth as a Pseudo-first-Order Reaction. mex—— THY ADDED AT £:=0 233py REMAINING IN SALT PHASE (%) ol | o0 apep AT r=%fl-fin1 | | | { - 10 —— b p— N 0 2 4 6 8 10 EXTRACTION TIME (hr) 7.9, this removal rate recasonably agreed with a first-order rate relation postulated for the reduction of protactinium by bpismuth droplets of con- stant thorium concentration while falling through the salt phase. The run was interrupted at this point to correct a malfunction in the pump. The apparent collection of solids about the pump rotor resulted in a seizure which could not be overcome by the low-torgue l/3-hp motor. A compilation of material balance data on thorium added to the sys- tem is shown in Table 7.7. Since lhe thorium baskets were coated heavily with bismuath, the actual quantity of thorium introduced ianto the system was estimated by volume loss rather than by weight balance. Spectro- graphic analyses of samples taken prior to each thorium addition showed that very little thorium or reduced lithium was retained as a soluble component of the metal phase. Chemical analyses of salt samples from - corresponding periods showed that the chromium concentration of the salt rhase was reduced from 164 to less than 2 ppm after the first thorium » addition and remained at that level for the duration of the experiment. Concentrations of iron and nickel were virtually unchanged at about 125 and 30 ppm respectively. During the interruption, the column of steel wool was removed from the system for examination. Although the column had gained considerable weight, only the head-end sections (the bottom of the column) contained appreciable gquantities of solids. A gquantitative analysis for 233Pa in the column was made by E. I. Wyatt, Analytical Chemistry Division, by dissolving the entire assembly 1into an aqueous solution. Radiochemical analyses of aliquots of this solution were related to the total volume cf solution and the quantity of 233pg, present originally in the salt sys- tem. On this basis, the column accounted for abhout 14% of the 237pa. Since 43% of the 222Pa activity remained in the salt and 3% remained in the metal, about 39% of the 233pg was apparently deposited elsewhere in the system. Table 7.7. Material Balance of Thorium Metal Added to #22Pa Extraction Loop Cumuletive Equivalence Additi Thorium Thorium horium Th Found Li Found Praction of Th ’ o0 ~otacted Withdrawn — Added 4As Th® As Th As 10 in Metal in Metal Accounted For No. () () (g) in Metal in Salt dn Metal (pom) (ppm) in Metal (ppm) {ppm) (ppm) 1 8.37 2.0 6.3 468 936 57 60 18 0.45 2 8.39 2 6.4 94,3 1836 114 30 13 0.13 3 5.9 ly oty 1.5 1055 2109 127 30 20 0.17 4 5.9 1.0 4.9 1419 2836 171 30 17 0.11 5 5.0 5.0 1790 3579 216 <30 17 0.0% £ot 154 The second steel wool column was assembled with coarser grades of steel wool in the inlet section in an attempt to retain more of the solids. In the previous column the filterable solids were stopped by the first three of eight sections of steel wool. It seems possible that solids held by inertia on the head end of the column settled in the pump bowl when the pump was stopped. The steel wool on the new column had a net weight of 18.2 g and a surface area of 0.87 m?. The results of radiochemical analyses of filtered samples from the two liquid phases taken during this second extraction experiment are summarized in Fig. 7.10. Two additions of thorium during 15 hr of pumping made no change in the 233pn concentration in the salt rhase. Because the pump was operating poorly and the flow rate of bismuth was irregular, subsequent thorium additions were made through the sampling port directly to the salt phase. Three l-g additions of thorium reduced the ?23Pa con- tent of the salt phase to about 27% of its original value before the run was Interrupted to change out the steel wool column. During this op- erational interval, 16% of the 22?Pa was removed from the salt phase, 4% was in the metal phase, and about 7% was found on the column. The steel wool column for the final extraction experiment was pat- terned arter the preceding one. A total of 28.7 g of steel wool from grades 3 through O provided a surface of 1.4 m?. Thorium metal was added in three 2-g increments during 12.5 hr of pump operation. Radiochemical analyses of filtered samples of the liquid phases are shown in Fig. 7.11. Protactinium was removed from the salt phase until about 4% of the 233pg remained at apparent equilibrium. During this period the pump was op- erated at its maximum allowable speed in an attempt to wash protactinium from the walls of the extraction vessel and to better suspend any prot- actinium-bearing solids in the molten bismuth. Material-balance calcu- lations show that 23% of the original quantity of 223Pa was removed from ORNL-DWG 56-11467 MOSTLY STATIC OPERATION WHILE PUMP // WAS SEIZED PRIOR TO 7=5 100 o I -t o @ ‘ | L | |3 = Q. g o - 1 80 |- - .J_ ] ;i{ . ..I__ ; S L; (2 &_ z 5 2y =i Fig. 7.10. Behavior of ?33pa j 2 L 5z e 1g. /.10, ehavior of Pa in —_ I < > - . -y - 2 o 2 %5; | g;; g Pump T,oop During Second Extraction 0 - o o A o 7 EI o . i 0 . 2 0 |- e E “fl*sfi | e gl Experiment (Th” Added to Bi Except o O 2 L S , i,e‘ ! Where Noted), > o i = A £ Y > ’ T l 3 i 0 o 2 <<'[) aon — o . L# o = L ac Q _J 0 SALT PHASE S 0 ® METAL PHASE | éa l s | - Ll i L o —es” r— - @ o L ‘ . . . . . ® N 2 0 5 1O 15 20 LOOP QPERATING TIVE (hr) o 155 ORNL—DW3 66--11468 O SALT PHASE 4o—§——~ ® METAL PHASE -— e e <5 g e Fig. 7.11. Behavior of ??°Pa o |3 i | , . . 5 30 e S I B — in Pump Loop During Third Extrac. @ tion Experiment (Th® Added to £ Salt Phase). 5 g 4 —-Q T —_ o oL rlgn. o ~ T ~ T 0 ( . [ 0 25 50 75 0.0 12.5 LOOP OPERATING TIME (hr) the salt phase. About 19% of the 23°Pa was on the column, giving an 82.6% recovery for this third experiment. Chemical analyses were obtalned on salt samples taken during the second and third extraction experiments. Concentrations of iron ranged randomly from 100 to 170 ppm, while chromium and nickel were virtually absent at reported values of less than 10 ppm. The nicbium content of the salt phase was consistently below the detectable level of 4 ppm. Values for the concentration of bismuth in the salt mixture ranged from 67 to 264 ppm in a random manner; the arithmetic average concentration from all samples was 119 ppm. However, it was not possible, under the experimental conditions, to ascertain whether these values represented dissolved bismuth or droplets suspended in the salt mixture. The values did not, however, reflect the pumping speed of the bismuth or a dependence on time. Discussion and Conclusions. The protactinium bhalance for the com- plete experiment shows that 26% of the 233pg, originally in the salt phase was removed by thorium metal additions. At least 43% was pumped as 8 so- Jution or suspension in molten bismuth and deposited on rather small volumes of steel wool. Since an additional 4% of the 232pa remained in each of the two ligquid phases, we can account for about 51% of the prot- actinium. Although large amounts of thorium were added to the system, the con- centration in bismuth remained very low. Considerable thorium had to be added in the first experiment bvefore any protactinium was extracted from thie salt. Additional thorium was consumed without extraction of prot- actinium at the beginning of the second experiment, which followed the changeout of the steel wool column. Reduction of some metal Tlucride impurities from the salt was evident, but the amount was not sufficient 156 to account for all the thorium losses. BHome high-melting metallic plugs were taken from the system, and the spectrograpnic analyses showed them to contain high concentrations of thorium asscocliated with iron and chromium. The collection of protactinium on the steel wool columns appeared to be primarily a filtration process, although some surface absorption also was apparent. We expected some mass transfer of iron in the poly- thermal loop system, but it appears that the transfer was aggravated by the presence of thorium. We believe that iron, chromium, and thorium, accompanied by some of the protactinium, formed compounds of low solu- bility in bismuth at the operating temperature. Precipitates formed, some of them collected on the filter, but the remainder collected in other parts of the system to account for the thorium and protactinium losses. This suggests that a more resistant material, such as niobium, would be desirable for use in the extraction equipment. The pump-locp experiments wlll be continued. The next one 1s planned to demonstrate recovery of protactinlium from the liquid-metal stream by contacting the bismuth with a molten salt that is saturated with a mixture of HF and Hp. Protactinium Studies in the High-Alpha Molten-3Salt Iaboratory — C. J. Barton An experiment on the removal of protactinium from a breeder-blanket mixture LiF-Th¥, (73-27 mole %) having an initial concentration of 25 ppm of 231pg was described in the previous report.27 Reduction of prot- actinium was eflected by metallic thorium in the presence of lead at about 625°C. The fact that only a small fraction of the reduced prot- actinium was found in the liquid lead encouraged study of other reduction techniques. Reduction with 3o0lid Tnorium. Several experiments were performed to study the reduction of protactinium in IiF-ThF, (73-27 mole %) by solid thorium in the form of a rod or turnings. Three different container ma- terials were used: nickel, copper, and graphite. The first experiment wags conducted in a nickel container with a 3/8-in.-0D thorium rod. The 231ps content of the salt mixture, based on analysis of filtered samples, dropped from 11.1 to 0.09 mg during the initial 65~min exposure of the rod to the molten fluoride mixture at 625°C. A further 5-hr exposure at the same temperature produced an apparent increase in 2°*Pa content to 0.54 mg. A large fraction {(approximately 70%, or 28 g) of the part of the rod that was immersed in the molten-salt mixture was lost during the experiment. We believe that the thorium rod was in contact with the bottom of the nickel pot during this experiment, causing a current flow that corroded the rod electrolytically. The salt mixture removed from the nickel pot contained a large amount of black material, part of which was magnetic. The analysis of the magnetic part showed 45% Ni, 30% Th, and 0.015% ??1Pa. The black, nonmagnetic material contained 22% Ni, 50% Th, and 0.027% 23 pa. 157 A second experiment, performed under conditions described above ex- ceplt that care was exerted to avolid contact of the thorium rod with the nickel pot, gave more readily understandable results. The 2?1Pa concen- tration of a filtered sample of the salt dropped to 30% of the initial concentration (32 ppm) after a 1-hr exposure and to 19% after the second hour of exposure. A ground sample of the uvnfiltered salt removed from tne pot at the conclusion of the experiment had a higher concentration of 231ps than the initial filtered sample. This may have been due, in part, to the fact that the precipitated protactiniwum was redissolved by HF-H; treatment in 2 smaller volume of fused salt than was present at the be- ginning of the experiment because of the removal of a significant fraction (about 2.5%) of the salt in each filtered sample. The next experiment was similar to the previous one except that thorjum turnings supported in a nickel-plated copper screen were exposed to the melt for 65 min and the 2?1Pa concentration of filtered salt de- creased to 9% of the initial value. The screen came loose from its sup- porting rod when we attempted to remove it from the pot, and it remained in the molten mixture. Conseguently, 1t was necessary to dissolve the thorium metal by prolonged treatment with an HP-H, mixture in order to redissclve the precipitated protactinium. The 231ps concentration of a filtered sample was 92% of the initial concentration after a 170-min HE-H, treatment, and, after an additional 43-min treatment, an unfiltered semple showed a content equal to 96% of the initial value. In the fourth thorium reduction experiment, we agaln exposed the melt containing 12 ppm of 231pg o thorium turnings held in a nickel- plated copper screen and found that a 120-min exposure removed 97.5% of the protactinium from solution, as determined in a filtered sample. This time, however, we removed the basket and turnings from the melt, cooled to room temperature with a helium atmosphere in the pot, and disassembled the apparatus to determine the distribution of reduced protactinium. The recovered salt contained 51% of the amount of 231py present at the be- ginning of the experiment, the basket and contents had 20%, the wall had 7%, while the dip leg and magnetic material removed from the salt (plus particles produced in sawing through the nickel pot) each contained avout 1%. About 16% of the protactinlum was unaccounted for. Fxperiments very much like that deseribed in the previous paragraph were conducted in copper and graphlite containers to study the effect of container material on distribution of reduced protactinium. In both cases, approximately 60% of the recovered protactinium was found in the ground, unfiltered salt (and untreated with HF and H,), although only 5% of the initial protactinium concentration was present in the final i1- tered sample of reduced salt in the graphite container experiment and 29% in the copper container. It appears, therefore, that a large fraction of reduced protactinium remains suspended in the molten LiF-Thi', mixture regardless of the container material. Electrolytic Reduction of Protactinium in IiF-ThF, (73-27 Mole %). A series of exploratory experiments on the electrolytic reduction of protactinium in molten IiF-ThF, (73-27 mole %) has been conducted with a variety of electrode arrangements. None of these arrangements has been explored in detail, and the preliminary results obtained in some cases 158 ars more confusing than enlightening. Consequently, these experiments will be sumrarized very briefly. Graphite anode — nickel wvessel cathode — 3.0 v and 0.5 amp: The protactinium content decreased during the first hour of electrolysis and then increased during the second and third hours for reasons that are not all clear. Silver ancde — grapnite liner or nickel dip leg as cathode -~ 3.0 v and about 1 amp: About 20% reduction in 231ps concentration after elec- trolyzing for 2 hr. Thorium rod anode — thorium rod cathode — graphite liner — 3.0 v and about 1.6 amp: Removed 95% of 231pg during 70-min electrolysis (filtered sample), but 95% of the initial ??'Pa concentration was found in the unfiltered salt. Nickel rod immersed in bismuth connected to negative side of 6.0-v battery — nickel rod immersed In the molten fluoride connected to positive pole of battery — 5.0 v and about 2.7 amp: No reduction in protactinium content of salt, and only a trace amount was found in the bismuth. Another construction of the previous experiment with bismuth cathode (i.e., nickel dip leg contacting both salt and bismuth) — graphite anode contacting fluoride mixture only — graphite liner ~ 5.0 v and 0.5 amp: Again, there was no significant reduction in the protactinium content of the salt. About 0.02% of the protactinium was found in the last filtered sample of bismuth as compared to 0.4% in the unfiltered bismuth. The latter also contained 6.3% nickel. Conclusions. Protactinium dissclved in molten IiF-Th¥, can be re- duced to a form that does not pass through a sintered copper filter, but a large part of the precipitated protactinium remains suspended in the molten mixture. Exploratory experiments on electrolytic reduction of protactinium have not produced encouraging results to date, but we are continuing to pursue this approach to the protactinium removal problem because of its potential simplicity. 7.4 Radiation Chemistry Xenon Diffusion and Possible Formation of Cesium Carbide in an MSBR - C. F. Baes, Jr., and R. B. Evans IITI Previously, several investigators have considered the neutron polson- ing effect caused by diffusion of 135%e into the graphite moderator of a molten-galt reactor.29732 It is the present purpose to consider as well the effects of the cesium which is born within the graphite by decay of the various fission product xenon nuclides which have diffused there. In 159 particular, it is of interest to estimate whether or not sufficient con- centrations of cesium might occur to form (lemellar) cesium carbides, as by Cso(g) + nC(s) = Can(s) 5 and whether or not a sufficient amount of CsC, could be formed to damage the graphite in a full~scale MSBER. In an attempt to answer these questions, the partial pressure of cesium within the graphite void spaces was calculated using the following model and assumptions: 1. Diffusion of gaseous xenon into the graphite void spaces and dif- fusion of gaseous cesium out of the graphite were approximated as one dimensional; that is, the moderator was represented as a slab of graphite infinite in two dimensions, with a specified thickness (21), immersed in the fuel salt. 2. All cesium born in the graphite was assumed to be in the gaseous elemental form. Cesium born in the fuel salt or reaching the fuel salt by diffusion from the graphite was assumed to be oxidized to CsT and to remain in the salt. 3. Steady-state conditions were assumed. The resulting expressions and the parameters employed (which corre-~ spond approximately to the present MSBR reference design??) are summarized in Table 7.8. The steady-state partial pressure of each cesium nuclide (PCS) was a function of the depth inlo the graphite (x), the porosity of the graphite (e), the partial pressure of the parent xenon at the salt- graphite interface (P ), the diffusion cocefficients of cesium and xenon (D, assumed to be the Seme for both), and the appropriate decay constants (2) and neutron capture cross sections (o). The xenon partial pressure at the salt-graphite interface (P%P) was, in turn, a function of several terms: Y corresponds to the xenofiJproduction rate; S reflects the xenon loss from the fuel salt by decay, stripping, and burnup; G reflects dif- fusion of xenon into the graphite; and, finally, ' is a factor reflecting the effect of the film coefficient H at the salt-graphite interface. (This film coefficient is defined by the relationship for the flux JXP, _ A0 JXe - H(CXe CXe) ? wherein CXe is the concentration of the nueclide in the bulk of salt and C§e is the concentration in eguilibriwan with P§e terms 5, F, and G also appear in the 135%e poison Tactor, which 1s the ratio of the 13°Xe poison fraction under the conditions specified to the maximum possible poison fraction, approximately 0.005. at the interface.) The The rather cumbersome expresslions in Table 7.8 were evaluated with a computer. The effects of variations in the gas stripping rate (AfiT), O 160 Teble 7.8, Caleuwlation of Cs® Partial Pressure and of 1°3°Xe Poisoning Parametersa Assigned MSRE Values Values ¢ Average thermal-neutron flux, 7 % 10%% 1.5 x 1013 cm < sec” Cas 235U concentration in fuel, 2 X 1074 1.8 X 1074 moles/cm3 T Temperature of core, °K 873 873 Qi Henry's law distribution ccefficient 6000 6000 for xenon ViV, Ratio of total fuel volume to fuel 4,37 4.0 volume in flux A/Vc Ratioc of graphite area to fuel volume 2.0 25 in flux, em™t L Half thilckness of graphite slab, 0.5 2 cm € Poreosity of graphite 0.05 0.09 Ag Fraction of fuel stripped per 0.001~0.1 0.0004~0,002 ' second, sac”* D Effective diffusion coefficient for 1078074 1.3 X 1074 both Xe and Cs, cmz/sec H Film coefficient at salt-graphite 0.002-0.02 0.00045 interface, cm/sec Partial Pressure of Cs Nuclide in Craphite 0 ey\)(e PCs - PXe —M_*““~—_~"~E*3 (ECS _-EXe) [ Six fii (ZIFX)} ( ZBiL 2 DCS (BXe 505 Ei =1 e + 2 Partial Pressure of Xe Parent at 5alt-Graphite Surface PO = F"S’%“@ ) Y = ¥y U25C250RT 135y%e Poison Factor a = (VT/VC) (AXe * KST) T Oy ? T e (1350 = Mss) Gi 2By L 135 e oD, B (e *° 1) P, = —e—— g | o> e xe Jr3s¢ + 7\135 H F =1 + 7 SET 5 + F‘g (e Xe 4 l) 2P, L AD. B (e e _ 1) G = Xe Xe < 7 25 T © (¢ ¢ 1+ 1) a, . . ¥, 0, and A denote, as usual, the yield, neutron capture cross section, and decay constant of a glven nuclide. 1el Fig. 7.12. Calculated Steady- State Pressure of Cesium at Center of Graphite Slab as a Function of the Gas Stripping Rate (NST)’ the Diffusion Coefficient (D) and the Film Coefficient (H). Other condi- tions are specified in Table 5.8. The upper curves show the corre- sponding variation in the 1°5Xe roison factor. 35%e POISON FACTOR 4 } : CESIUM PRESSURE {atm ORNL — DWG 66-11468 H {cm /sec) / .02 ESTIMATED PRESSU o0 )‘ST (sec~1) o " fem?/sec) 106 RE AT WHICH CsCy IS FORMED the diffusion ccefficients (D and the film coefficient (H) were deter- 2 mined. Cesium Carbide Formation. The maximum total pressure of cesium (at the center of the graphite slab) was found to increase ama D was decreased (Fig. 7.12). This is a varadoxical result since it is usually thought desirable that the graphite possess the lowest possible D values. In the range of D values tested here, however, the rate~controlling step in the diffusion of parent xenon into the graphite appeared Lo he at the salt- graphite Iinterface (see below) and did not depend significantly upon D. 162 As a result, lowering D for xenon and cesium appreciably decreased only the diffusion of cesium out of the graphite, causing the steady-state accumulation of cesium_(PbS) to be higher. Included for comparison in ¥ig. 7.12 is the partial pressure of cesium at which carbide formation might be expected at 600°C, based on an estimate by Manowitz.?% Tt appears that in all cases tested the cesium partial pressure in the graphite was high enough to produce carpide for- mation. As a consequence, it is likely that the actual quantities of cesium which could accumulate within the graphite will be higher than estimated in the present calculations. While no detailed calculation of this effect was attempted, considerable comfort may be Tound in the Tol- lowing observations. 1. By the present calculations the amount of cesiwn accumalation in the graphite is very small; even with a xenon partial pressure as high as 0.01 atm, for example, there would be only 1 atom of cesium present for every 25,000,000 atoms of carbon. At acceptably low *7°Xe poison- ing levels, the calculated maximum cesium concentration would be ap- proximately 100-fold (10~% atm) lower than this. Thus, although the present estimates of cesium accumulation would probably be increased if carbide formation occurs, an enormous increase (e.g., 10°-fold) in the accumulation could occur before it would become a matter of concern. 2. Bven if 1t were assumed that all the cesium born in the graphite were to remain there, the rate of accumulation would be low. Thus, under the most unfavorable conditions tested (H = 0.02 cm/sec, A, = 0.001 ST Sec"l, and D = 1076 cmz/sec, which gave a 135Xe poison factor of 0.9), the total of all xenon nuclides entered the graphite at a rate v = JXe = E%K 5 GP%e = 1.29 x 107t mole cm™? sec™* 3 so low that even if it all were to decay to cesium, 150 days would be required to produce a cesium-to-carbon ratio of 1:1000. With more realistic calculations and a more realistic calculation of the geccumulation rate — in particular, a calculation which includes the diffusion of Can as well as gaseous cesium out of the slabh — much lower accumulation rates undoubtedly would result. 12°¥e Poisoning. For most of the values of D and H chosen here, the rate-determining step was the transfer of xenon across the film at the salt-graphite interface (i.e., F >> 1 and G/F — AH/V&QH). Hence the poison fraction was not very dependent upon D, but rather was deter- mined almost entirely by the magnitude of H and K%T (Figs. 7.13 and 7.14). This result, it should be noted, 1s compatible with the assumption of one=- dimensional diffusion in the graphite; that is, 1if the rate step instead had been in the graphite, such a crude model would have been more question- able. 163 ORNL-DWG 86-11470 H {cm /sec) ! o 0.02 0.0+ 0.005 0.002 05 0.2 r o © a £ O ? S 0 Q) > ©w " 0.05 0.02 0.04 1078 07 108 1073 10? 0 (cm®/sec) Fig. 7.13. Dependence of the Xenon Poison Factor on the Diffusion Coefficient. The 35%e polson factor, as well as the cesium partial pressures, could of course be reduced either by increasing KOT or decreasing H (Fig. v 7.14); however, it seems unlikely that in practice the stripping rate could greatly exceed 0.01 (corresponding to a removal half-time of ap- proximately 1 min) or that H would normally be kept below 0.01 cm/sec. Approximately these conditions evidently must be met 1f the poison factor is to be held below 0.1 (poison fraction approximately 0.005). ORNL-DWG 66-11471 OO‘F """" "“T """" CTTT T 0005 | i j{ —. 0002 |- ‘{ O & MAXIMUM 'O xe £ POISON FACTOR = ~ 0.001 L e 00005 | — ] ’ 0.0002 |— T —_t L1 0.0001 i e L I 0001 0002 0005 0.01 0.02 0.05 04 Agr (sec™!) Fig., 7.14., Approximabte Boundary Values of The Stripping Rate and the Film Coefliclent Calculated for Various Xenon Poisoning Limits in the Absence of Todine Stripping. Todine Removal. Removal of iodine (as HI) by HF sparging of the salts would be an effective way of reducing the 135%e poison fraction, provided the overall removal rate is large compared with the decay rate of *3°T (2.89 x 107% sec™, t1;p = 6.7 hr) — and it seems likely that this could be done.?? The amount of the reduction in the case of <220 fission- ing could be as much as & factor of 5, limited by the 18 * 3% of the “*°Xe being produced directly Trom fission. This treatment is relatively ineffective in reducing the cesium partial pressure, however, because the principal contributors to it are mass numbers 137 and 138, which have short-lived iodine precursors (A = 0.03 and 0.12 sec™ respectively). Salt Impregnation. Tapregnation of the graphite to a limited depth with the fuel salt would be an effective means to reduce the rate of xerion diffusion into the graphite by virtue of the fact that both the xenon diffusion coefficient and the xenon concentration would be roughly four corders of magnitude lower in an interstitial salt phase than in an interstitial gas phase. This situation can be repreéesented within the framework cof the present calculations by the simple procedure of considering the interpenetration- salt—graphite barrier as the film to which H refers. Comparison of the equation which defines H in a normal film, J =1 (-~-c?), 165 with the equation which would define the diffusion coefficient in the salt- graphite barrier, Y- shows that 1t is reasonable to approximate H for such a barrier by H~ D/t , where © is the depth of salt penetration. The effective D for this salt- graphite barrier should be in the range 1077 to 108 cmz/sec for MSRE- type graphite (i.e., D for xenon in the molten salt times the porosity/ tortuosity factor). Assuming a penetration depth of 0.1 cm, we can then place H in the range 107° to 1077, which is at least three orders of magnitude below the minimum film factor estimated for an MSBR. The consequences of such a small film coefficient are that diffusion of xenon into the graphite is no longer significant in determining the 135%e poison factor (i.e., the terms containing F in the expression for the poison factor become negligible). The 133ye poisoning is then deter- mined only by %ST: 0135® "o (VT/VC) (M35 + ?\ST) + 01350 P.F Summary. The present calculations indicate that cesium carbide Tor- mation can be expected to occur in an MSBR, but in such small amounts as to be of little concern. In addition, these calculations indicate that in the absence of iodine removal, xenon poisoning in a full-scale MSBR will he controlled primarily by a film coefficient H and will be reduced most effectively either by iodine removal or by some method which in ef- fect reduces this film coefficient. Fission Product Behavior in the MSRE — S. S. Kirslis The behavior of fission products in the M3SRE is being studied to derive information bearing on the practical concerns of corrosion and neutron poisoning. Significant clues to the basic nature of Hastelloy N corrosion in a fissioning molten-salt environment are provided by obser- vation of the volatilization and plating behavior of fission products whose oxidation states are relatively easily changed. The chemical fate of a number of fission product poisons 1s important in determining their contributions to the overall polsoning of a reactor. Thus, Tor example, it is of interest to determine what fraction of the T3%Xe produced by fission in the melt 1s released to the cover gas before 1t becomes 136%e by neutron capture. Iikewise, it is 1mportant to determine whether noble-metal fissicon products which are moderate neutron poisons (Mo, Ru, Te, Nb, Pd, and Rh) remain in the circulating fuel, volatilize, or deposit 166 on graphite or Hastelloy N. The rare-carth fission product poisons, be- cause of the chemical stability of their nonvolatile fluorides, are ex- pected to remalin in the circulating lfuel. There arsc unique advantages in carrying out fission product studies in an operating reactor rather than in small-scale laboratory or in-pile tests. Iirst, in small-scale tests it 1is difficult to mock up realisti- cally the geometry, the fuel circulation conditions, and other possibly significant factors of the reactor environment. Second, although the MSRE was not primarily designed for the chemist's convenience, several features make it rather well adapted to chemical studies. Most importantly, a facility exists for taking sizable samples of fuel salt from the pump bowl during reactor operation. In the same facility 1t is possible to expose selected metal sampies to both the cover gas and the fuel melt in the punp cowl. Provisions exist for taking samples of the reactor cover gas and are being improved to permit continuous monitoring and the taking of concentrated samples. Finally, provisions werec made for Inserting re- movable long-term surveillance specimens of Hastelloy N and graphite in the reactor core of the MSRE. These facllities and advantages are avail- able only with great difficulty in small-scale in-pile tests. The prin- cipal lacks from the chemist's viewpoint are facilities for direct sam- pling of the pump bowl cover gas and for exposing metal and graphite sam-~ ples to -the pump bowl atmosphere and fuel phase for longer periods of time. This report on fission product behavior will cover the initial re- sults from fuel salt sampling, from the exposure of metal samples to the pump bowl gas and liquid phases, from the first examination of the long- term survelllance specimens of graphite and Hastelloy N, and from the analysis of the first sample of reactor cover gas taken during reactor operation. Ffuel Salt bamples. Using the pump bowl sampling facility, a large number of 10- and 50-g fuel salt samples were taken during reactor op- eration to monitor changes in the concentrations of oxide, bulk fuel components, and corrosion products. Metal samples were attached to the 10~g sampling assembly for observation of the volatilization and plating behavior of Tission products during five of the salt samplings. These five salt samples were delivered to the analytical hot cells within a few hours of sampling and were rapidly powdered, weighed, dissolved, and analyzed radiochemically for the 15 isotopes listed in Table 7.9. The strontium and cerium isotopes were delermined as fission monitors, since these elements have convenient half-lives and stable fluorides which ghould remain in the melt. The ruthenium and molybdenum isotopes were representative of noble metals. The tellurium and icdine isotopes were of interest for their volatilization and plating properties as precursors of 13%Xe. The 239Np and *29Pu isotopes were measured as indicators of the epithermal fiux in the MSRE. The data listed in Table 7.9 show that the strontium and cerium isotopes behaved in a regular and expected manner. The °lor and *43ce isotopes appear to be the most reliable fission monitors for interrupted power operat.ion. The average of the four values for °lsr at the nominal Table 7.,9. TFission Products in MSRE Salt Samples Sarple TPo=-17 P6-19 FP7-7 FP7-10 FP7-12 Date and sampling time 5-23, 040 5-26, 0400 6=27, 0243 7-6, 0208 7-13, 0336 Operating time,a days 2.0 2.5 13.3 el 1.9 Nominal power, Mw 5.6 7.3 ey 7.2 7.2 Fission . Yield Disintegrations per Minute per Gram of Sal1t° (%) 9.67-ar “iSe 5.81 1.08 x 10+ 1,20 x 0% 1.16 x 10t 1,31 x 10t 1.32 x 10t 2.6-hr °%gr 5.3 9.8 x 10 1,19 x 0% 9,70 x 1019 1.51 x 10*? 1.49 X 10+t 51-day 2%r 4,79 1.80 X 10%0 2,23 x 10*° 2.96 x 0% 2.93 x 100 3.96 x 10%9 33-day *4iCe 6.0 2,65 % 1040 6.10 % 10*° 6.69 x 10*° 6.88 % 10%° 33-hr *43Ce 5.7 1.45 x 0% 1.5 x 10 1,43 x 10t 1.32 x 10%* 66-nr ?Mo 6.C6 .68 X 10%9 3,51 x 10 2 9,51 x 10*9 1,10 x 10* 3.15 x 10%9 39,7-day *%%Ru 3.0 1.81 x 0% 7,05 x 10° 2.42 X 107 6.02 x 107 7.14 X 107 lidB-hr *O9Ry 0.9 4.9 X 10%% 2,47 x 10** 2 3.50 x 100 9.67 x 10%° 3.76 % 1049 1.0L-year 19%Ry 0.38 0 2.13 x 10" 77-nr 1327 AR 3,12 ¥ 10F0 4,21 % 10%° 5.15 x 16*% 3,64 x 10%° 3,81 X 10%° 8.05-day 13T 3.1 2.45 X 100 4,20 x 100 5,0 x 10Y0 4,50 x 1010 5.36 x 1049 20,8-hr 1331 6.9 1.08 x 10% 1,31 x 10+t 1.35 X 10** 1.40 x 10** 1.45 x 10 6.7-nr *3°1 5.1 9.5 x 10*% 1,50 x 0%t 1.17 X 10%* 1.11 x 10tt 2,33-day 231 5,35 X 10% 1.0 x 102 1.08 x 10%2 2.44 X 10%-year Pu 7,00 X 107 ¢ 8,77 x 10° ¢ .20 x 10% ¢ a . . X . . - . e s . Continucus cperating time since previcus shutdown of more than 12 hr duratlon or change in power level. b Calculated at sampling time. C . Alpha counts per minute per gram., LOT 168 Table 7.10. Bummary cof Pump Bowl Test Results Averages of five runs 99MO lSZTe lO'SRU. 1033ua lOéRu 135I lBBI 1311 Salt, % of theoretical 60 30 >1.00 30 15 90 1.00 98 Lateh (Ni) 8 14X 10X 10X — 0.5X 6X 02X 1.5% Silver 2X 6X 3% 3% — 1X 2% 01X 0.9x% Hastelloy N 1% 7 3% 25X — 0.3% 11X 0 2x 0.5% Liguid phase (ss) 4K X 5% 40X — 1X 1X 0 2X 0.8X Spercent 2Ry deposited on metal samples uniformly decreased from first to fifth run. bX = disintegrations per minute of given isctope per gram of salt. 7.2-Mw power level is 1.25 x 101 dis min™?t g™*. Using the formula Mw = [(1,25 x 10 dis min"l ¢71) % (4.68 x 10° g) L x (200 Mev/fission) x (1.6 x 10719 . M )J 3+ [(fractional ev/sec fission yield of ?lgr) x (60 sec/min)] , the calculated fission power density is 5.4 Mw, or about 75% of the nominal total power density. Using the average of the four 14306 values at the nominal 7.2-Mw power, the calculated power is 6.3 Mw, or 87% of the nominal value. - For molybdenum and ruthenium isotopes, the concentrations found in salt samples showed much more scatter, which may well be real. The first row of data in Table 7.10 shows the averages of the amounts of these isotopes Tound compared to the amounts calculated to be formed using 91gy (arbitrarily) as the [ission monitor in each run. The reason for the impossibly hign values for 20°%Ru is being sought. The 1327¢ values varied little among the 7.2-Mw runs but represented low fractions of the total amount formed (Table 7.10). The iodine isotopes showed relatively little scatter, and the amounts found corresponded well with the amounts calculated from the 2'Sr values. The epithermal flux in the MSRE has not yet been calculated from the neptunium and plutonium values in Table 7.9. Pump Bowl Volstilization and Plating Tests. Advantage was taken of the experimental possibilities of the pump bowl salt sampling facility to carry out some gualitative tests to detect the presence of chemically reactive fission product species in the pump bowl cover gas and to deter- mine which fission products would plate on clean metal surfaces from the fuel salt phase. 169 The copper fuel salt sampling ladle 1s attached by a double stain- less steel cable to a nickel-plated irvon latch which fits into a mecha- nism for lowering and raising the assembly in a pipe leading from the pump bowl to the sampling cubicle above the reactor. To detecli the presence of chemically reactive fisgion products in the gas phase, colls of 0.015-in.-diam silver and Hastelloy N wire were wound on the small sbainless steel cables fTor a distance of 2 1n. below Lhe bottom of the latech. The latch and the wire colls were later leached and analyzed for gagsaous Tisslon products. The lower 2~in. lengths of the stainless steel cables Just above the copper ladle were leached and analyzed similarly to determine which fissgion products plated from the fuel melt. Figure 7.15 shows the sampling assembly with wire coils attached. The gampling device was submerged in the pump bowl for times verying from 1 min to 10 min. It was thsn-ralo d 2 ft and allowed to cool for 10 min and then ralised to the sampling cublcle. The metal and salt samples were delivered in & carrier to the hot analytlcal laboratory, usually within 3 hr of the sampling time. The salt sample was prepared for analysis by the procedure summarized above. The stainless steel cables were clipped to provide separate samples of the lateh, the silver coll, the Hasztelloy N coil, and the stainless steel cable exposed to the fuel melt. The metal samples were leached Tilrst with an alkaline mixture of Versene, boric acid, and citric acid to remove iodine without volatilization. Then they were leached with a warm mixture of HNO3 and HC1 until the radiocactivity was less than 1% of the original reading {(ususlly greater than 500 r/hr at contact). Dilutions of the original leach solutions were sent to the radiochemical separations group for overall gamma scans and estimations of’ several lfldLVldUdl Tzotopes. The overall gamma scans showed that the principal sctivities in the metal samples exposed to the pump bowl gas phass were 1227 and ]321 with smaller amounts of 21T and ??Mo. The *4PRa-*40ra and 2%7r- 95Nb peaks, which were prominent in the gamma, spectra of fuel zalt samples, were much lower in the metal samples. This indicates that the observed activities were not due to condensation of fusl salt mist on the metal specimens. Similarly, the gamma spectra of the submerged stainless steel samples were quite different from those of Tuel salt. Correspondingly, no adhering particles of fuel salt were visible on any of the mebal speci- mens. It was later found {Table 7.10) that the amounts of tellurium, rutheninm, and molybdenum deposited on the metal samples corresponded to the amounts in srfgrdW grams of fuel salt. The gamma scans indicated that quantitative radiochemical estimations for *2%0e, T and Y%Mo would ve useful. In addition, estimations were made of 1331 1357, leRu 105Ru) and °%Ru. The results of these analy- BC8 are oummdrized in Table 7.10. Although the exposure times for the five runs were 1, 2, 5, 10, and 10 min, there was no corresponding vari- ation of amounts of the several isotopes deposited from either the gas or the Jiguld phases. The scatlter of res 1t3 for a given isotope between runs was large, often a factor of 2 and sometimes a factor of 4 or more. For these reasons, the averaged results given in Table 7.10 are more casily interpreted than =z tabulation of the individual results. As In the case of some of Lhe salt analyses, the oObserved scatter is much greater than the normal error of radicchemical analyses and undoubtedly 170 PHOTO 85018 0 1 .. " Fig. 7.15. Pump Bowl Deposition Testing Assembly. represents real large variations in the amounts deposited. The causes for these variaticns are as yet unknown. The results in Table 7.10 reveal a striking degree of volatilization and plating of molybdenum, ruthenium, and tellurium on clean metal sur- faces. Iarge amounts of égMo were found on the metal samples in both the gas phase and fuel phase of the pump bowl. Correspondingly, only about 60% of the theoretical yield of ?°Mo was found in the fuel salt 171 samples. The results for ruthenium were similarly spectacular. The fractions of the three ruthenium isotopes remaining in the salt decreased with increasing half-1life, suggesting that slow reactions are removing ruthenium from the melt. In successive runs, the amount of 41-day 1°3Ru that deposited on metals steadily decreased from the first run to the last (although the exposure time of the samples increased from 1 to 10 min), as if the amount in solution were decreasing with time. This behavior is not explained. For 122Te the amount remaining in the melt is also low (30%), and the corresponding gas- and liquid-phase depositions are high. For 1311, 1331, and 13511 most of the element remains in the melt, although high activities of 1321 were observed on the gas-phase specimens. The short- lived 221 is, however, a daughter of the relatively long-lived l32Te, which was shown to volatilize readily. Rather less 1337 and 1317 was found on the metal specimens than of molybdenum, ruthenium, and tellurium. However, no 1351 at all was detected on the metal specimens. Corre- spondingly, the tellurium precursor of 1357 has a half-1life of only 0.4 min, while 133Te and *?!Te have half-lives of 63 and 24 min. Apparently very little tellurium can volatilize or plate in 0.4 min. It thus appears that the volatilization and plating characteristics of iodine are deter- mined by the behavior of the precursor tellurium. Tellurium and iodine had been detected in the gas lines of previous in-pile tests, but the volatilization of molybdenum and ruthenium was surprising. The only volatile compounds of molybdenum and ruthenium are those of valence 4 or greater. The plating behavior of tellurium, mo- lybdenum, and ruthenium also indicates that these elements are present with valences greater than zero. However, all positively charged ions of these elements should be reduced to the metallic state by the metallic chromium in Hastelloy N. These considerations suggest that the Hastelloy N container vessel and piping of the MSRE may be protected by a layer of noble-metal fission products, permitting the existence of higher oxida- tion states in the melt. If it is assumed that 30% of the noble metals produced by fission in the MSRE operating at 7.2 Mw are deposited on an estimated metal surface area of 10° cm?, it may be calculated that the metal film would grow at the rate of about 0.3 A/hr. It is thus not un- likely that all the metal surfaces exposed to fuel in the MSRE are presently coated with several hundred angstroms of noble metals. It is not a simple matter, however, to produce adherent pinhole-free electroplates, so that a plate hundreds of atoms thick might not protect against the reducing action of the base metal. Furthermore, 1t has been calculated on a very reasonable basis that approximately 1% of the uranium in the fuel in the MSRE should exist in the trivalent condition. Noble- metal ion concentrations should be infinitesimal in the presence of this concentration of the strongly reducing U3+, These very basic puzzles in elucidating noble-metal fission product behavicor In the MSRE have not yet been solved. Future pump bowl experiments will involve looking at the behavior of several more elements (noble metals, other cations with volatile fluorides, neutron-activated Hastelloy N corrosion products, etc.) and attempts to 172 take small samples of the pump bowl cover gas. The determination of the reducing power of the fuel salt would also be helpful in explaining fis- sion nioduct behavior. Examination of the Graphite Survelllance Specimens. A package of MSRE graphite and Hastelloy N surveillance specimens has been exposed to a fissioning molten-salt environment along the central axis of the MSRE for 7800 Mwhr of power operation. After the reactor shutdown of July 17, 1966, the package was removed from the reactor and disassembled in a hot cell. Rectangular bars of graphite, 5 to 9 in. long, 0.66 in. wide, and 0.47 in. thick, from the bottom (inlet), middle, and top (outlet) of the reactor core were made available for detailed examination. The surveil- lance specimens were contained in a perforated cylindrical tube of Hastelloy N, 5-1/2 ft long, 2 in. in diameter, and 0.030 in. thick. Rings of this tube approximately 11/16 in. in height and 10 g in weight were . sawed out of the bottom, middle, and top regions of the tube to provide Hastelloy N specimens for fission product deposition studies. - The graphite bars were first sectioned transversely with a thin Carborundum saw to provide specimens for photographic, metallographic, autoradiographic, x-radiographic, and surface x-ray examination. Speci- mens were saved for other possible tests (spark spectroscopy or electron probe). The remainders of the bars, 7 in. long for the middle specimen, 2-5/8 in. long for the bottom sample, and 2-7/8 in. long for the top sam- ple, were used for milling off successive surface layers for fission product deposition studies. (a) Visual, Autoradiographic, and X-Radiographic Examinations. The graphite specimens showed no visible signs of corrosion or chemical change. No metallic or salt films were discernible under low-power mag- nification, which revealed the original machining grooves on the graphilte. The middle sample (Y-2) was known to have a crack near one end before in- sertion into the reactor. The crack was still visible after removal but had not propagated further. Some of the graphite specimens in one region of the package had cracked due to mechanical stresses caused by the un- even thermal expansion and contraction of the tightly packed Hastelloy N and graphite specimens. A small piece of the bottom sample (VA-1) had cracked off during the disassembly of the package. These cracks should not reflect on the overall integrity of the irradiated graphite speci- mens. Prints of an autoradiograph and an x radiograph of each of the three graphite specimens are shown in Figs. 7.16 to 7.18. The x radio- graphs were taken on thin (0.020- to 0.060-in.) transverse slices of the graphite bars. The autoradiographs were taken on adjacent transverse slices after mounting in epoxy resin and polishing for metallographic examination. The x radiographs show no general deposition of fuel salt nor of heavy elements on the surface of the graphite. A hint of penetration or deposition is visible near one corner each of the bottom and top graphite samples. The crack in the middle sample is clearly penetrated by fuel salt. The edge of the cracked bottom sample shows no foreign material, indicating that the crack occurred after removal from the reactor. (a) ig. 7.16., Autoradiograph (a) and X Radiograph (b) of Middle CGraphite (v-2). The autoradiographs indicate a thin film of highly radicactive ma- terial on the exposed surfaces of the graphite and confirm the presence of salt in the crack of the middle specimen. These pictures also show that the penetration of the radlioactive material into the interior of the graphite is by no means unliform. The nonuniform nature of the porosity of graphite has been demonstrated in gas permeability tests. The observations described here are generally in accord with those from previous in-plle tests. [The graphite is not visibly affected, and there are no signs of chemical attack, film formation, or salt genetration. (b) Metallographic Examination. Metallographs of transverse sections of the three graphite bars are shown in Figs. 7.19 to 7.21. The stiructure of the graphite in all cases appeared normal and undamaged under both bright-~-field and polarized-light illumination. No metallic, carbide, or sall films were visible on the surfaces of the specimens. No sign of salt penetration was observed except in the case of the cracked middle speci- men. Here voids were oObserved which probably had been filled with salt before polishing with a water suspension of the polishing compound. The other graphite specimens were polished under CCl, to avoid dissolution of fuel salt. 7.18. s S 0 .&‘.&%"‘ R &‘s*::'% i "t O o e : ,_,&,:?:,—j_“‘g. o . : S e 2 : fi%fi%‘#@ : i R : oI - e e . ! S '-‘.;.% Autoradiograph (a) and X Radiograph (b) of Botbtom Graphite 175 O ™| X008 o« SAHINI LCOC sl X00§ = SR OO0 ) L] :, /M —~ @ o § b ~ @ 2 o g £ o & O w — T ke i 0 0 w0 = a2y 3 . & £ O el — —l @ LT M J o ~ + @ oy~ —~ O - By o~ P D o By en T l T T AR R o Tha e e X¢ y 05 o LO0C R G e e, t field; ) Brigh a ( 20. Metallographs of Top Graphite {(VH zed light 7 Fig. (b) polar i i 177 D.007 INCHES EQOX = CHzS ox J 0.007 1k n Fig. 7.21. Metallograpns of Bottom Graphi . . 4 . phite (VA-1). o field; (b) polarized light. ( ) (a) Bright 178 A fuel-salt-exposed surface of the middle graphite bar was mounted for hot-cell x-ray examination of the surface for metallic films or other contamination. Only graphite lines were observed. Because of the nega- tive result of this test, the other graphite specimens were nol examined by the x-ray technique. (¢) Milling of Surface layers of Graphite. The valusble contribu- tions of J. G. Morgan, M. F. Osborne, and H. E. Robertson in planning, developing hot-cell methods for, and starting the work on the sampling of the graphite specimens are gratefully acknowledged. A very ingenious "planer" was designed and built by the Hot Cells Operation Group for mill- ing thin layers from the four long surfaces of each of the graphite bars. The cutter and the collection system were so designed that a large fraction of the graphite dust removed was collected. By comparing the collected weights of the graphite samples with the weight loss calculated from the initial and final dimensions of the graphite bars and their known den- sities, the average sampling losses were 4.5% for the middle bar, 18.9% for the top bar, and 9.1% for the bottom bar. The patterns of sampling of the graphite surface layers are shown in Fig. 7.22. An identifying groove was cut along the length of the middle of the graphite surface which was pressed against another grapnite sur- face in the original surveillance package. The other ftThree surfaces were exposed to circulating Tuel salt. The numbers of the layers in Fig. 7.22 indicate the order in which the layers were cut from the graphite bars. After each layer was cut from a bar, the milling apparatus and the bar were vacuumed to avoid cross contamination of samples. Rach powdered graphite sample was placed in a small capped plastic bottle and weighed. The middle bar was measured with a micrometer to determline the depth of each cut. This time-consuming operation was omitted for the other two bars, since it became evident that the depth of cut could be more accurately calculated from the weight of the sample removed. The average depth of cut was 00,0075 in. for the middle bar and 0.011 in. for the other two bars. A clean unirradiated MSRE graphite specimen was sampled with the milling device in the standard manner after the ninth cut and after the last cut on the middle graphite bar to provide comparison sam- ples to indicate the level of cross contamination in the hot-cell milling operation. As seen in Fig. 7.22, the salt-exposed surfaces of the middle bar were sampled to a depth of six or seven layers, or about 0.050 in. The surfaces in contact with graphite were more thoroughly sampled in the other bars. (d) Fission Product Analyses in Graphite Samples. The powdered graphite samples were delivered to the Radiochemical Separations Group, and weighed portions were dissolved in a hot mixture of HNO; and ,50,. The gases evolved were passed through a condenser, a charcoal trap, and an alkaline solution to recover any volatilized iodine or ruthenium. Aliquots of these solutions were analyzed radiochemically Tor the isotopes listed in Table 7.11. The first milled sample of each graphite surface was also analyzed for uraniuwa by the fluorometric method. No uranium was detected with a sensitivity limit of about 30 ug of uranium per square centimeter of graphite surface. This finding will be confirmed by more sensitive neutron actlvation methods. 179 ORNL-DWG 66-11378 TOP GRAPHITE e IDENTIFYING GROOVE O O Mepleo]es MIDDLE. GRAPHITE - 0470 in, | | .| BOTTOM GRAPHITE A Fig. 7.22, Scheme for Milling Graphite Samples. The radiochemical data so far reported are given in Tables 7.11 to 7.13 for the middle, top, and bottom graphite bars respectively. The data are generally internally consistent for a given bar, and the agree- ment between bars is good. Very few individual values are out of line, which is to the credit of those involved in the sampling and in the anal- ysis of the samples. From the neutron poison staandpoint, the results of most interest are those for Mo, *°2Ru, and ?°Nb. It is seen ia Tables 7.12 to 7.13 that the concentrations of these isotopes are considerable in the first layer and that the activities fall off by a factor of about 100 in the seccnd layer. Tellurium-132, which is also noble in the sense of possessing fluorides of only moderate stability, behaves rather similarly, but its concentration drops off less rapidly with penetration distancs. DBeyond a penetraticn depth of about four cuts, the activities of molybdenum, 180 ruthenium, and nicbium approach the contamination background (samples 10 and 24 in Table 7.11), while that of 1327 is distinctly higher. It is pregsently thought that the deposition of molybdenwn, ruthenium, and proba- bly niobium in the graphite may be due to the reaction of volatile hexa- fluorides or pentafluorides with graphlite, depositing lower fluorides or carbides. There is no chemical reason 1o expect the lower fluorides of molybdenum or ruthenium to plate on graphite. On the other hand, the pump bowl experiments definitely established the volatility of molybdenum and ruthenium, probably in the forms of MoFg and Rufs. The behavior of l4oBa, SQSr, lélCe, 14408, and 13705, all of which have xenon or krypton precursors, is distinctly different. The gradient of activity with penetration distance is much less, and it appears to bear a relationship to the half-1ife of the rare-gas precursor involved. Thus the gradient for 14084 with a 16-sec 140%e precursor is much steeper than that for 89Sr, which has a 3.2-min %Kr precursor. Apparently, the longer-halif-1life rare gases can achieve a Tlatter gradient by diffusion before they decay to the observed isotope, which is assumed to remain where deposited. The penetration data for 1%YBa and 8%Sr fitted well a simple diffusion model which led To Xenon and krypton diffusion coeffi- cients of about 1 x 107 ftz/hr in MSRE graphite. To explain why 13708, which has a 3.8-min '?7Xe precursor, has a much flatter profile than 8%sr in Table 7.11, it may further be postulated that ?7Cs itself diffuses, rather than remaining where it was born. The internal *?7Cs concentration of 2 x 107 dis min™* g%, or about 1 atom of *27Cs per 108 atoms of graph- ite, should have negligible chemical effect on the integrity of the graph- ite structure. Only a few values have yet been obtained for %7r. ‘This element is expected to remain in the fuel salt, with little tendency to volatilize or plate. Thus the amount found in the graphite should repre- sent only injection by fission recoil. Correspondingly, the amount in the graphite is low compared to the noble metals. The concentration of T2'I in the graphite is alsc low at the surface, and its §radient is similar to that of *??Te. It is possible that the 24-min 21Te diffused into the graphite as TeFg or TeF, and then decayed to AL, However, it is difficult to see why the iodine did not diffuse vack out of the graphite, unless 1t formed a nonvolatile compound with other fission preoduct atoms in the graphite. While it is not possible at this time to account satisfactorily for the Tact that the ilodine concen- trations in graphite are low, the implication of the low 1317 concentra- tions is that the 12°I concentrations will be similarly low. Thus little 13%%e will be born in graphite due to the previous immigration of 1357, Since the depths of cut of the many graphite samples were not the sane, especially for different bars, it is not casy to compare the amounts of the various isotopes in the first layer from the data in Tables 7.11 to 7.13. BSince, also, the main practical interest is in the noble-metal fission products, the bulk of which were deposited in the first layer, Table 7.14 was prepared, giving the amount of each isotope in the first layer in disintegrations per minute per square centimeter of graphite surface. The values in Table 7.14 were obtained by multiplying the values in Tavles 7.11 to 7.13 (disintegrations per minute per gram) by the weight of sample analyzed (corrected for weight loss in sampling) and dividing by the measured area of the surface from which the sample was cut. The 181 Table 7.11. Radiochemical Analyses of Middle Graphite Bar (Y-2) Weight Depth of Disintegrations per Minute per Gram of Graphite Sample Cut , (e mits) 990 132.0, 1035, 95\ 1314 95 144, 89g, 140, 1410, 137 VWide Face Exposed to Circulating Fuel 1 0.8463 6.02 1.88 x 1012 1.09 x10*2 2.57 x 10! 6.93 x10!' 128 x10°° 1.12 x10'? 5,08 x10° 128 x10'! 144 x10't 3.7 x10'® 8.39 x107 4 1.2737 9.27 1.28 x10*°% 5.19 % 10'? 2.40 x10° 9.16 x 10° 9.45 x 108 8.43 x 107 1.10 x10'! 6.57 x10'? 6.53 x10° 2,02 X 107 7 0.9814 7.50 2.35 x 10° 2.33 x 1019 4.24 x 108 7.76 X 10° 4.17 x10° 2.80 X107 7.65 x10'% 2.75 x10*% 1.43 x10° 1.98 % 107 11 0.9145 6.94 5.53 x 108 1.07 x10'° 1.52 x 10% 2.94 x 108 1,77 x 108 2.22 x 107 7.08 x10'% 1.65 x10'° 7.33 x 108 2.17 x 107 14 0.8962 6.98 1.20 x 109 8.60 x 107 3.17 x 108 4.28 % 10° 3.18 x 108 4.07 X107 4.55 x10'% 9,71 x10° 5.55 % 108 1.86 x 107 17 1.0372 8.25 1.13 x 10° 8.27 x 10° 2.18 x 10° 3.16 x 107 1.79 x 108 2.46 X 10° 430 x10'% 7.60 x10° 3.90 x 108 1.96 x 107 23 0.8176 6.64 1.12 x 10° 8.76 x 10° 1.73 x 10% 2.43 x 10% 7.74 % 10° Side Face Exposed to Circulating Fuel 2 0.7583 7.68 1.40 x10'? 1.04 x10'? 2.14 x10'! 833 x10'' 1,20 x10'° 2.48 x 107 1.41 x10** 3.38 x10'° 6 0.9720 10.10 2.70x10'° 462 x10'? s5.92 x10® 152 x10'% 8.98 x10° 7.68 x 10'° 8 0.5139 5.43 3.73x10° 2,41 x10'% 6.33 x 108 2.79 x 10° 3.00 x 107 3.86 x 10 2,50 x10° 12 0.3395 3.65 2.44x10'° 379 x10'% 4.55 x10° 1.36 x 109 2.99 x 1010 15 0.6976 7.61 466x10” 111 x10'% 7.38 x10° 2.75 x 10% 2.54 x 107 1.52 x10'% 8.39 x 10° 18 0.3737 4.15 2,41 x10'% 305 x10'% 4.76 x10° 6.62 % 10° 2.29 x 10" Other Side Face Exposed to Circulating Fue! 3 0.6135 6.21 1.73 x 1012 1.09 x10'? 2.60 x10'' 6.36 x10'" 1.61 x10'7 2.45 x 10° 1.75 x 10" 3.35 x10'° 5 2.7543 7.84 5.94x10'° 7.01 x10'® 8.55x10° 2.80 x10'% 1.21 x10? 7.18 x 1010 9 0.6198 6.55 853x10° 196 x10'® 5.52 x10° 1.79 x 10% 2.18 x 10’ 4.28 x10*° 3.01 x10° 13 1.1652 12.53 6.26 x10° 870 x10% 1.15 x10° 2.05 x 10° 1.00 x 10%7 16 0.8469 9.24 561 x10° 8.02x10° 7.25%10° 1.41 x 108 2.04 x 107 4.98 x10° 3.77 x10° 19 0.9406 10.45 5.08 x10° 5.65x10° 6.89 x 10° 6.99 x 107 4.23 x10° Face in Contact with Graphite 20 1.1381 9.25 3.22x10'! s5.82x10'" 7.88x10'% 104 x10'' 1.36 x10'? 6.68 x10° 1.65 x10° 8.96 x10*° 2.15 x10'° Unexposed Graphite Blanks 10 4.63 % 10° 2.62 x 10% 7.18 x 107 2,46 % 10° 3.26 % 107 4.48 % 10° 1.07 x10° 2.03 x10° Low 1.41 x 10° 24 6.47 x10° 3.22x10® 811 x107 328 x10° 126 x10° 1.31 x10° 8.06 x10° 3.21 x107 6.38 x10° NOTES: 1. The samples are arranged in order of successive cuts on each face (see Fig, 7.22). 2. 'The sample weights given here have been corrected for the average 4.5% loss during milling, 3. The depths of cut were calculated from the sample weights, areas, and the known graphite density. 4. Additional analytical results are forthcoming on selected samples for 9SNb, 89Sr, QQTC, 952r, 147Nc§, 136Cs, 137Cs, 91Y, °3Ni, and ?Fe. 5. The activities tabulated are corrected to the time of shutdown, 11:00 AM, Jfuly 17, 1965, 182 Table 7,12, Radiochemical Analyses of Top Graphite Bar (VH.5) Disintegrations per Minute per Gram of Graphite Sample Weight Depth of {g) Cut {mils) %90 132m 1055, 95N 1314 957, 1440, 89g, 1405, t4la, 1374 Wide Foce Exposed to Circulating Fuei 25 0.3602 6.23 1.54 x 1012 @89 x10°! g.02x10'! is6x 10! 4.87x10° 1.28x10% 440x10® 1.19x 10!t 1.04x10'' 860x10° 2.24 x 107 29 0.4355 7.93 5.45x 107 9.24 x 10° 6.30 x 10° 1.06 x 10° 4.92 x 107 5.74x 101% 1.60 x 161° 1.22 x 107 58 0.5260 9.94 5.95 x 10° .39 x 10° 7.76 x 108 9.54 x 103 6.54 x 107 6.19x 10'? 1.98x10'® 1.87 x10% 60 0.2916 5.51 1.73 x 10° 3.95 x 10° 1.64 x 108 3.58 x 107 3.95x 101" 7.19 x 10° 62 1.0783 20.38 6.22 x 10° 1.63 x 107 7.96 x 167 1.34 x 107 1.5: x 1089 1.67 x 10° 3.54 x 107 Side Foce Exposed to Circulating Fue! 26 0.4615 11.41 4.94 x 1017 438 x10'! 1.37x10%! 2.37% 10% 2.59 x 10% 4.96 x 10'® 4,94 x 10° 31 0.4564 11.88 2.24 x 10° 9.06 x 10° 2.82 x 108 7.26 x 107 6.11 x 108 1.45x 10" 2,13 x 108 Other Side Face Exposed to Circulating Fuel 28 0.6703 17.11 6.10 x 10'1 5.7t x10%' 7.81 x 10'° 5.44 x 107 2.69 % 108 6.38 x 10'% 5,38 x 10° 33 0.5404 14.43 2.20 x 107 5.39 x 10° 2.32 x 108 3.01 x 107 8.27 x 10° Wide Face in Contact with Graphite 27 0.6422 11.32 3.86 x 10*Y 2,80 x 10'Y 514 x10'% ss50x10't 3.07x107 1.24x10° 2.74 x 10® 3.29 x 1010 4,03 x 10° 32 0.5375 9.95 1.956 x 107 1.03 x 10'° 1.91 x 10 2.28 x 108 9.26 X 107 8.51 x 10° 59 0.3154 5.96 1.81 x 107 1.24 x 10*® 2,01 x 10° 6.25 x 108 7.31 x 107 8.75 x 10° 2.56 x 10° 61 0.5835 11.03 9.17 x 108 3.95 x 107 1.17 x 10° 1.86 x 107 2.76 x 107 63 0.7310 13.82 6,08 x 10% 2.41 x 10° 8.16 x 107 1.1t x 107 1.58 x 10° 2.17 x 107 NOTES: The samples are arranged in order of successive cuts on each face (see Fig. 7.22). Additional analytical results are forthcoming on selected samples for QSNb, 898r, The sample weights given have been corrected for the average 18.9% weight loss during milling. The depths of cuf were calculated from the sample weights, areas, and the known graphite density. 99TC' 957, 1475a, 138cs 137cs 91y S3Ni and S9Fe. The activities tabulated are corrected to the time of shutdown, 11:00 AM, July 17, 1966, 183 Table 7.13. Radiochemical Anciyses of Bottom Graphite Bar {VA-1) Weight Depth of Disintegrations per Minute per Gram of Graphite Sample - Cut ; (g) (mns) QQMO 132Te 103Ru QSNb ].EH.1 QSZr 144Ce SQSr 140Ba I41Ce 137CS Wide Face Exposed to Circulating Fuel 34 0.8032 15.04 5.30 x 102! 4.39 x10'1 8.30 x10'% 3.34 x10'Y 4.25 x10° 2.40 % 10° 6.39 x10% 4.17 x10'? 3.34 x10'% 8.11 x107 2.80 X 107 38 0.5979 11.64 5.67 x 10° 3.26 x10'? 8.22 x10® 2.49 x 10° 4.44 x 108 3.86 x10'? 1.65 x 1019 1.09 x 107 64 0.2323 4.68 9.5 x10° 1.91 x10'% 9.76 x 107 3.12 x io® 2.63 x 108 3.52 x10'% 1.05 x10'° 6.92 x10° 56 0.3120 6.28 8.65 x 10°% 1.91 x10*% 1.20 x 10% 1.40 x 10% 3.62 x 1019 1.13 x 100 69 0.7183 14.49 3.39 x 108 1.06 x10*? 5.14 x 107 8.01 x 107 3.08 x10!% 5.87 x10° 1.84 x 10°% Side Face Exposed to Circulating Fuel 35 0.3904 10.70 6.62 x10'* 5.31 x 10! 9.44 x10'° 7.09 x 10° 8.32 x 10° 4.18 x10'% 1.04 x 1019 39 0.4480 12.98 7.56 x 10° 4.12 x101% s5.86 x 10° 5.44 x 108 2.21 x 1029 Other Side Face Exposed to Circulating Fuel 37 0.5480 15.39 4.3% x 10'Y 3.64 x 10! 5.11 %3010 4.58 x 10° 6.19 x 10° 6.12 x101% 9.19 x10° 41 0.3520 9.62 3.95 x 107 1.81 x10'% 4,95 x 108 1.43 x 108 1.38 x 109 7.25 x 10% Wide Face in Contact with Graphite 36 0.4810 9,12 0.31 x 10! 5.97 x10'' 1.00 x10'' 4.20 x10'! 6.01 x10° 3.47 % 10° 9.49 x 108 4.48 x10'% 1.07 x10'° 40 0.5936 11.77 8.23 X 10° 3.70 x10'% 1.06 x 10° 2,50 % 10° 5,01 x 108 2.04 x 1019 65 0.4756 9.58 1.57 x 10° 2.44 x16° 1.74 x 108 3.07 x 10% 4.31 x 108 1.48 x10*? 1,10 x 108 67 0.4025 8.10 1.83 x 10° 2.15 x10'% 1.48 x 108 1.31 x 108 1.24 x 101¢ 68 0.6260 12.61 2.66 x 108 6.90 x 10 3.83 x 107 4.96 x 107 4,50 x 10° 1.03 x 108 NOTES: 1., The samples are arranged in order of successive cuts on each face (see Fig. 7.22). 2. The sample weights given have been corrected for the average 9.1% weight loss during milling, 3. The depths of cut were calculated from the sample weights, areas, and the known graphite density. 4, Additional analytical resulits are forthcoming on selected samples for QSNb, 898:‘, 99Tc, 95Zr, 147Nd, 136Cs, 13'7Cs, le, 63Ni, and SgFe. 5. The activities tabulated are corrected to the time of shutdown, 11:00 AM, July 17, 1966. 184 Table 7.14. Fission Product Deposition on Surface® of MSRE Graphite Top Graphite Middle Graphite Disintegrations per Disintegrations per Bottom Graphite Disintegrations per fsotope Minute per Square Percent of Total® Minute per Square Percent of Total? Minute per Square Percent of Total® Centimeter Centimeter Centimeter 9910 3.97 x 1010 13.36 5.14 x 1017 17.24 3.42 x 1010 11.5 1327 3.22 x 101° 13.84 3.26 x 10*° 13.60 2.78 x 10'° 12.0 193gu 8.34 x 107 11.40 7.53 x 10° 10.32 4.75 % 10° 6.30 Nb 4,62 x 109 12.00 2.28% 101? 59.2 2.40 x 10°° 62.4 131 2.00 x 108 0.162 4.22 x 108 0.328 3.27 x 108 0.252 Szr 3.78 x 107 0.326 3.14 x 108 0.270 1.72 x 108 0.148 e 1.58 x 107 0.0516 8.26 x 107 0.268 4.36 x 107 0.142 89gr 3.52 x 107 3.24 3.58 x 10° 3.30 2.99 x 10° 2.74 140p, 3.55 x 10° 1.38 4.76 x 10° 1.85 2.93 % 167 1.14 4ice 3.16 x 108 0.194 1.03 x 10° 0,632 5.83 x 10° 0.356 137¢s 6.62 x 10° 0.070 2.35 x 108 0.248 2.01 x 10° 0.212 60(."10 in adjacent 4.37 x 10° 1.57 x 109 6.69 x 10° Hastelloy N, a i i dis min~ " g -1 a u " s Average on the three salt-exposed surfaces for first milied cut. Ppercent of total amount of isctope in reactor system which is deposited in the 2 x 106 cm? of graphite surface that is in the reactor. 185 deta were condensed by averaging the usually well-agreeing amounts Found on the three salt-exposed surfaces. The amounts found on the other graphite-contacting surfaces were, on the average, 38% lower. The latter figure indicates rather free circulation of salt between the contacting graphite surfaces. Oalt may be expected to circulate egually well in the interstices between the graphite stringers of the MSRE. It is seen from Table 7.14 that the amounts of noble metals in the top, middle, and bottom samples of the graphite agree rather well, with pernaps usually somewhat larger values for the middle graphite. For the other isotopes, also, the middle graphite usually holds the highest sur- face concentrations, though seldom in the ratio of the prevailing neutron fluxes. The last row in Table 7.14 gives the °°Co activities in the Hastelloy N samples adjacent to the three graphite samples. These numbers are proportional to the thermal-neutron fluxes. The latter have not yet been calculated since the chemical cobalt analyses have not yet been re- ceilved. Also listed in Table 7.14 are the percentages of the total isotope produced by Tission found on the graphite surface, on the assumption that 511 2 x 10° cm? of graphite held the surface concentrations listed to their left in the table. The total amcunt of each isotoove was calculatad for a fission rate based on the °'gr activity (1.32 X 101+ dis min~t =) in the last salt sample taken on July 13, 1966. The percentages in Table 7.14 would be sbout 20% lower if the nominal power bistory of the reactor were used to calculate the total amount of each isotope produced by fis- sion. It should be recognized that the percentages glven actually repre- sent the behavior of each izotope only during a few half-lives of that isotope. Thus the figures for 66-hr “?Mo charscterize only a period of a week or so before the July 17, 1966, shutdown. It is quilte conceivable that the deposition behavior of the various isotopes is changing with time. Several sets of survelllance specimens must be examined before [irm conclusions may be drawn. Sizable percentages of the total r”9I\/Io, 132Te, lOBRu, and 2°Nb were found on the graphite surfaces. Many of the values were close to 13%, but two of the ?7Nb values were near 60%. Much lower percentages of the other isotopes deposited on the graphite, the highest being 3.2% for 893y and 1.8% tor T49pa,. An attempt vwag made to znalyze two of the surflace graphite samples for all fission product isotopes of molyovdenum by mass spectrometbry, after spiking the samples with °2M0 or 28Mo. The results indicabed somewhat more molybdenum deposition (20 and 32% of the total) than was measured radiochemically with “?Mo. However, confidence in thesze resulbs was lessened by the fact that distinetly different combinations of fission product molybdenum, natural molybdenum, and natural zirconium (used as a carrier) had to be used to fit the mass analyses of the two samples. (e) Fission Product Deposition on Hastelloy N. It is interesting to compare the deposition of fission products on graphite with that on the adjacent Hastelloy N. The samples of the perforated Hastelloy N ob- tained as described above were weighed and dlssolved in a warm mixture of HNG3z and HC1L. From the known dimensions and density of Hastelloy N, it was calculated that the surface-to-welght ratio of the perforated metal Table 7.15. Deposition of Fission Produets on Hastelloy N Top Middle Disintegrations per Minute Percent of Hastelloy N per Square Total® Graphite Centimeter o 2.12 % 10%! 42,8 5.3 1326 5.08 x 10°} 131 15.8 19304 3,55 % 10 © 29.3 4.3 13y 8.24 » 107 3.82 39,4 957, 1.85 % 10° 0.96 48.9 lag 5.22 % 107 0.019 .14 144, 1.07 x 107 0.020 0.68 60Co in adjacent 9 4,37 < 10 Hastelloy N, :: o—1 =1 dis min g Disintegrations per Minute per Square Centimeter 2.76 % 10" ° 3.41 x 1011 2.55 x 190 2.7 x 107 1.84 x 10° 2.24 x 10° 8.95 % 107 Percent of Tota 55.6 88 1.84 0.95 0.07 0. 17 1.57 x 10 Bastelloy N 1 Graphite 5.3 i.4 3.4 9.4 5.9 i 0.22 5 1.1 10 per Minute per Square Centimeter 2.04 % 101 4.27 % 10"} 2.32% 100 5.24 % 10° 2.5% % 107 @ 1.50 % 107 3.51 % 107 Disiniegrations Bottom Percent of Hasteiioy N Total Craphize 41.2 5.9 110 15.3 19 4.9 2.44 5.0 1.32 15.0 0.055 0.25 0.068 9,81 6.69 x 10 .. , . . . : o . . . 2 . : : . ; APercent of total isotope produced by fission in the reactor which was deposited on 1.2 % 107 cm” of Hastelloy N surface in the reactor, assuming deposition on all surfaces is the same as on Hastelioy N in the core. hRatio of the dis min_l em™?% of each isctope on Hastelloy N to that in the first milled cut of the adjacent graphite. 981 187 was 4.30 em®/g. Using this figure the anslytical results (in dis min~?* o~ %) could be converted to dis min™t em™%. Table 7.15 gives the radio- chemical results obtained to date from the solutions of the Hastelloy N specimens. It is seen that large percentages of the total 2“Mo, 132Te, and 103Ry deposited on the Hastelloy N. Also given in Table 7.15 are the ratios of the surface concentrations of each isotope on the Hastelloy N and on the graphite. Between 3 and 14 tTimes as nuch molyodenum, tellurium, end ruthenium deposited on metal as on graphite. It Is in- teresting that a falr material balance is obtained for 29Mo by adding the vercentages of the total produced found in the salt (approximately 60%, Table 7.10), in the graphite (approximately 14%, Table 7.14), and on the Hastelloy N (approximately 45%, Table 7.15). The agreement for the other isotopes is not so good. Miuch lower percentages of the total 1311 deposit on the Hastelloy N, but the amount compared to that on gravhite i1s high. The data suggest that the precursor tellurium deposits on the metal and that not all of the 1317 is able to leave when the tellurium decays. The amount of 237 on the metal was low, butbt higher than the surface concentration on graph- ite. Very little 14iCe and f440@ were found on the metal, in fact, less than on graphite. The fioding of more cevium isotopez in graphite may be due to their short-lived xenon precursors. General Discussion of the Deposition Results. The principal prsctical interest 1n the fission product deposition results lies in thelr implica- tions regarding neutron poiscning by noble metals in the graphite cores of molten-salt reactors. In most studiez of the physics of reactors such as the M3BR, it has been assumed that the ncble-metal fission products would either plate out instantanecusly on metal surfaces Or e removed periodically from the fluoride fuel by the Tluoride volatility orocessing. Under these conditions, the poisoning effect of noble metals would be of minor practical concern. However, it has been calculated®® that if all of the elements 3e, Br, No, Mo, Te, Te, and I remained in the core of the - MSBR, the neutron poiscon fraction (capture by these elements per absorption in fissile material) would be 0.0969 in two years, 0.1652 in five years, 0.2104 in ten years, and 0.2585 at equilibrium. IT 10% of these materials remainad in the core, the respective poison fractions would be 0.0106, 0.0172, 0.0226, and 0.0275. If this group of slements remained in the fuel and was periodically removed by fiuoride volatlility processing, the average poison fraction would remain constant at 0.0015. The isotopes of mo lybdenum and 9%Te gecount for about 90% of the polsoning effect of this group of elements. The nuclear breeding ratio for the MEER is expected to be 1.05 to 1.07 in the absence of deposits of noble metals in the core. The numbers zmbove indicate that the actual breeding ratio or the time that the graphite can be left in the core of a breeder can be considerably influenced by the deposition of this group of elements. 1t was also reported above that five to six times more 29Mo was de- posited on Hastelloy N surfaces than on adjacent grapnite surfaces. In the MSRE there is approximately 2 X 1 % em?® of graphite surface (about half in the fuel channels and hall in the flats pressing against adjacent flats) and 1.2 x 10°% cm? of Hastelloy N surface exposed to the circulating fuel. In the current MSBR design the ratio of metal surface to graphite 188 surface exposed to circulating fuel is about 2. It seems likely that the fraction of molybdenurm depcsited on the gravhite in the MSBR core would be correspondingly lower than in the MSERE. The ratio of metal to graphitce surface arca coula be further increased, but at the expense of fuel in- ventory, by circulating the fuel through a chamber packed with finely divided metal. Anotner approach to the problem of molybdenum deposition in graphite is suggested by considerations of the probable chemistry involved. The volatility of 290 demonstrated in the pump bowl tests suggests the presence of Morg Or MoF; in the circulating Tuel. The gaseous form of molybdenum would explain 1ts observed penctration to appreciable deptiis into the graphite surfacoe. On the basis of available thermcchemical in- formation, it is not clear what chemical reacticn occurs to fix the molybdenum to the graphite. For example, the free energies of the reac- ticns of Molbg with graphite to form CI, and molybdenum or Mol, are posi- tlve by a foew kilocalories. The same is true for Rul's and NbFs. Only in the case of tellurium is there a definltely negative free-energy change for the reaction of the [luoride with graphlte. Possible chemical ex- planations of the observed depositions are the formation of stable mixed alloys of the noble-metal fission produclts or the reactions of the [luo- rides with reducing impuritics present in the grapnite. In any case, these reactions take place inside the graphite pores near the surface. The penetration of volatile molybdenum could be decreased by making the graphite surface less permeable. 1t is hoped that grapghite speclmens whose surfaces have been made less permeable by impregnation treatments can be included in futurc surveillance specimen packages to test whether molybdenum deposlltion in graphite can te decreascd in this way. It the deposition of noble-metal Tission proaucts in grapnite does indeed reguire thelr prior conversion to high-valent [luorides, a third method of deposition control suggests itself: +the fuel melt may be made more reducing by increasing the U3+/U4+ ratio. Fxisting thermochemical data indicate that a sizabkle percentage of the tetravalent uranium could be converted to the UPF state without causing uranium carbide formation. A small percentage of the urarium could be made trivalent by adding a few percent of Hz to the helium cover gas of a reactor. Alternatively, when fresh fuel is added to compeansate for burnup, the added uranium could be partly trivalent. One or both of these procedures may be Teasible in the MSRE. The effects on volatilization and plating behavior could be easily observed by pump bowl tests with metal and graphite specimens. The second major area in wnich illumination is desired from fission product behavior is Hastelloy N corrosion. As indicated above, 1t 1is difficult to reconcile all the observations with a single consistent chemical pvicture. The principal difficulty is that the high-valent forms of molybdenum, ruthenium, and tellurium indicated by thelr volatility and plating behavior may not exist in equilibrium with the appreciable con- centrations of U’ estimated to be present in the MERE fuel salt. The oxidizing effect of burning up about 400 g of ?3°U to date i1s equivalent to only one-tenth of the UeT originally present in the Iuel. If it is assumed that the U’V concentration was actually mucen lower than estimated, the presence of high-valent noble metals in the fuel im- plies that the Hastelloy N surfaces have been protectively plated with 189 noble metals. Experimental observations of such platings were reported apove. This protection would explain the observed low corrosion rate of Hastelloy N as indicated by chromium analyses of the fuel. However, the absence of iodine velatilization is then difficult to explain. In the presence of MoFg, RuFs, and Telg, it is expected that icdides would be oxidized to Ipz. It is not expected that the solubility of I, in the fuel is high encugh to prevent its velatilization. These problems of consistency make it clear that the nature of Hastelloy N corrosion reactions in an operating reactor is not under- stood as well as would be desirable. Of great help in solving some of these problems would be the development of a method to measure the Ut concentration in radloactive fuel salt samples. MSRE Cover-Gas Analyses. Samples of MSRE cover gas were Ilsolated on June 23, 1966, during steady 7.2-Mw operation in the three shielded gas-sampling bombs provided for this purpcse. The three samples (500, 500, and 1500 em’ ) represented pump bowl cover gas which had passed through the first holdup volume (68 ft of 4-in. stainless steel pipe), the particle trap, and the charcoal filter. The latter should have re~ moved all heavy hydrocarbons and other impurities more easily adsorbed than xenon. The gas analysis of prime interest was that for the l36Xe/134Xe ratio, from which the fractional burnup of 135%e in the MSRE could be determined. To obtain high-sensitivity mass spectrometric analyses for 136ye and 134Xe, it was necessary to concentrate the sampled gas by a factor of at least 50. The concentration was accomplished by adsorbing the impurities in a helium cover-gas sample on a small volume of molecular sieve sorbent at liguid-nitrogen temperature, then warming the sorbent €O 500°F, and flushing the liberated impurities with helium into & small (20- cem®) sample bottle. In this way, one 500-cm? sample was concentrated by a factor of 25 and the 1500-cm® sample by a factor of 75. At the same time small unconcentrated samples of the cover gas were taken for gamma spectrometry (1 em?) and for mass spectrometry (2 cmB) to detect impuri- ties such as H;0 which are not readily liberated from the warmed molecular gsieve sorbent. The latter sample also provided more reliable analyses for very low-boiling impurities like Hp, which would not be completely sorbed by the molecular sleve. After cooling for more than two months, the radicactivity of the gas samples was low, and no activity problems were encountered in the direct gampling and concentration procedures. The activity in the gamma spectro- metric sample, taken in a thin-walled l-cm? glass bulb, was preponderantly S.27-day 1334, The observed counting rate indicated a concentration of about 7 ppm of 133%e in the cover gas at the time of sampling. Minute tracez were also detected of a2 0.16-Mev activity which might be 12-day 13Myve and of a 0.5-Mev activity which might be 10.3-year 85Kr, 40-Aday 103ry, or 1.0-year 106Ry. The results of the mass analyses of the unconcentrated sample, the concentrated 500-cm’ ssmple, and the concentrated 1500-cm> sample are shown in Table 7.16. From the ratio of the analyses for 136xe ang **%Xe for the concentrated 300-cm? sample, it was calculated that 7.7% of the 120 +35%e produced in the fuel melt captured neubrons to become *30Xe. The corresponding more accurate result from the 1500-cm® sample was 7.9% T35%e burnup, with a standard deviation of about 0.5% caleculated from the esti- mated analytical accuracy. It should be emphasized that this method of determining 1355 oburnup 1s not affected by sample contamination or by the exact value of the concentration factor, provided the concentrated Table 7.16. Mass Spectrometric Analyses of MSRE Cover Gas Sample No. OG-8 0G-5 oG-7 Concentration factor 1 25 75 Constituent, P Hy C.022 0.81 0.49 He 99.71 93.65 S5.27 CH, < 0.005 0.44 0.23 H>0 < 0.005 0.058 0.021 Hydrocarbons 0.005 0.010 0.004 N, + CO 0.1 3.87 2,70 Oz 0.066 0.49 0.26 Ar 0.003 0.043 0.028 CO» 0.006 0.53 Q.54 Kr < 0.005 0.018 0.083 Xe < Q.01 0.078 0.38 Isotopic Analyses Sample No. Theor. 0G-5 OG-6 Constituent, % 83yr 14.1 14.0 14.05 84y 25.93 26.8 26.52 85gy 7.60 7.6 7.4 8oy 52.37 51.6 51.96 131ye 13.42 8.8 9.08 132ve 20.06 14,7 14.72 134%e 36.92 41.1 40.90 136ye 29 .59 35,4 35.30 - 191 seample contains sufficient xenon for a good measurement of isotople ratios. The low +2°Xe burnup values are consistent with indications from MSRE re- activity measurements and are much lower than values predicted theoreti- cally for the case of no helium bubbles circulating with the fuel salt through the MSRE core. Recent measurements indicated an appreciable bubble volume fracticon in the circulating fuel. From Table 7.16 it is seen that the observed fractions of 1?'Xe and 132%e were much lower than the fractions calculated from the flssion yields. In the case of 131Xe, this is well acccunted for by the fact that the precursor &.05-day 1317 had not yet reached its equilibrium concentra- tion when the sample was taken. For 132Xe, the observed value was some- what lower than is indicated by a similar explanation involving the 77~hr 132me precursor. The observed krypton isotopic percentages matched closely those calculated from fission yields (Table 7.16). The observed ratios of total xenon to total krypton were lower than the theoretical 5.7, proba- bly because of the xenon precursor effects mentioned above. It may be noted that the total xenon and total krypton concentrations were a factor of nearly 5 higher for the 1500-cm® sample than for the 500-cm? sample, whereas a factor of 3 was expected. The concentrations of all impurities in the three analyzed samples should have been in the ratiocs of their respective concentration factors. This was generally not true. Except for total xenon and total krypton, the impurity concentration in the 500-cm” sample was considerably more than the expected one-third of the 1500-cm® sample concentration. Most analyses for impurities are therefore uncertain by a factor of 2 or more. The analyses for air (Np, 0,, and Ar) were higher than expected for the thoroughly leak-checked sampling and concentration system. A pure helium sample is being concentrated in the system to provide a quantitative plank for the alr values. The analyses indicate more than 60 ppm of Hp and COp and more than 30 pp of CH, in the MSRE off-gas. Presumably, these gases were generated from the thermal and radiolytic decomposition of the organic materlals which caused plugging in the off-gas system. The low concentrations of nigher hydrocarbons indlcate that the particle trap and charcosl filter were still effective on June 23, 19606. 7.5 Development and Evalvation of Analytical Methods for Molten-Salt Reactors Determination of Oxide in Radiozctive MSRE Samples — R. F. Apple, J. M. Dale, and A. &. Meyer The equipment for the determination of oxide?” in highly radiocactive MSRE salts has been transferred to the High-level Alpha Radiation Labo- ratory and installed therein. Since installation, one coolant salt and 192 - Table 7.17. Oxide Concentrations of Coolant and PFuel Salt from the MSRE 1 oncentration Sample Code Oxide Conce o1 (ppm) Coolant salt CP-4-4 25 Tuel salt I'p-6-1 £9 FP~6-4 53 FP-6~12 50 - FP-6-18 457 FP-7-5 66 FP-7-9 59 . rp-7-13 66 _ FP-7~16 56 eight fuel samples have been analyzed for oxide. The 50-g fuel salt sam- ples were taken at levels from zero- to full-power reactor operations. With an in-cell radiation monitor, the initial sample read 30 r at 1 ft. This activity increased to 1000 r at 1 £t at the full-power level. Re- sults of the oxide analyses are given in Table 7.17. The oxide 1n the coolant sglt sample, 25 ppm, is comparable to a value of 38 ppm obtained for a coolant salt sample taken on January 25, 1966, and analyzed in the laboratory after three weeks' storage. The fuel analyses are 1n reasonable agreement with the samples analyzed on the bench top before the reactor was operated at power. The oxide in these nonradioactive samples gradually decreased from 106 to 65 ppm. Between the FP-6 and FP-7 series the sampler-enricher station was opened for maintenance, and the apparent increase in oxide concentration (ca. 15 ppm) may represent contamination of the samples by residual moisture in the sampling system; however, the number of determinations has not been suf- - ficient to establish the precision of the method at these low concentra- tion levels. In an attempt to determine whether radioclytic fluorine is removing oxide from the fuel samplies, sample FP-7-9 was removed from the transport container and stored in a desicecator for 24 hr prior to analysis. Since the oxide content, 59 ppm, is comparable to that of the remaining samples for which analyses were started 6 to 10 hr after sampling, no significant loss of oxygen 1s indicated. A more direct method of establishing the validity of these results by measuring the recovery of a standard addition of oxide will be attempted when reactor operations are resumed. A method for verifying the performance of the capillary gas stream splitter and the water electrolysis cell in the remocte oxide apparatus was developed. A tin capsule containing a known amount of 5n0O, 15 heated to 550°C in the hydrofluorinator as hydrogen is passed through the system. e 193 The SnO, is reduced to the metal, and the water formed passes on to the electrolysis cell. Two standard samples of S5n0,; were analyzed with a four-month interim, and oxide recoveries of 96.1% and 95.6% were obtained. It is probable that this slight negative bias is due to momentary interruptions in the flow of the hydrofluorinator effluent gas through the water electrolysis cell. Difficulty with cell plugging was en- countered throughout the period of development of the oxide method. As an attempt to eliminate the negative bias and also to provide a replace- ment cell for the remote oxide apparatus, it was deemed necessary to find & method of regenerating the electrolysis cell which would permit a steady gas flow at relatively low flow rates. The water electrolysis cell contains partially hydrated P05 in the form of & thin viscous film in contact with two spirally wound 5-mil rhodium electrode wires. The wires are retained on the inside of an inert plastic tube forming a 20-mil capillary through which the sample passes. The 2-ft-long tubing element is colled in a helix inside of a 5/8—in.-diam pipe and potted in plastic for permanence. During the course of the investigation of the cell, 1t was found that a wet gas stream in iteelf did not cause the electrolysis cell to plug. It was also necessary for current to be flowing through the cell for flow interruptions to cccur. This indicated that the hydrogen and oxygen evolving from the electrodes create bubbles In the partially hydrated P05 film, which then grow in size sufficient to bridge the capillary and form an obstructing film. After many unsuccessful approaches, an acceptable answer to the problem was obtained by means of a special regenerating technique using dilute acetone solutions of H3PO, as the regenerating solutions. This provides a desiccant coating sufficient to absorv the water in the gas stream and gives a minimal amount of flow interruptions during electrol- ysis. The cells which have been successiully regenerated in this manner have yielded oxide recoveries of 99.6 * 1.3% from standard Sn0, samples. Spectrophotometric Studies of Molten-5alt Reactor fuels — J. P. Young Studies pertaining to a continuous spectrophotometric determination of U(IIT) in circulating MSBR fuels have continued. A general description of the optical design of a facility for performing this determination has been discussed.’® The facility will be used in conjunction with a com- mercial instrument, a Cary model 14H recording spectrophotometer. During this period further and more detalled discussions of the optical problems involved have been carried out with the designer of this spectrophotometer and have culminated in a purchase order issued to them for the development of suitable sample-space optics. The apparatus which will result from this requisition is to be delivered in six months and will provide optimum light gathering power and optical design for use with a double convex lens~-shaped drop of liguid. The apparatus will be used with existing in- strumentation and will be compatible with the optical path extension which willl be required in the final proposed reactor installation. 194 More detailed studies of the spectra of U(III) in molten 2IiF-Bef, have provided better resolution of the various absorptions in the complex ultraviolet absorption peaks of U(III). As reported before, the maximum abvsorption is 360 nm, but shoulders are observed at approximately 310, 445, and 508 nm. These shoulders might be of analytieal value if as yet unknown interferences vprevent the use of the absorption at 360 nm. On the basis of caleculations made to estimate the probable valence state cf Tission products, it is expected that dissolved fission products will be in one of their more normal valence states and therefore will cause no interference at the concentrations expected. Concerning unusual oxldation states of rare-ecarth fission products, cursory spectral studies have been carried out with Sm{IIl} in molten 2LiF*Be¥,. This lower valence state of samariwa exhibits a strong, broad absorption peak with a maximum absorption at 325 nm with a shoulder at approximately 470 na. The molar absorptivity is not yet known, butl it is believed to be greater than 200. TIf conditions are such that Sm(II) is rresent, possible interference with a determination of U(IIT) at 360 nm may be encountered, depending on con- centrations present. Interference of lower-valent rare earths will not be a general problem, but only a problem with specific ions; for example, Fu(Tl) exhibits nc absorbance at wavelengths above 300 nm. Although it has been assumed from other specltral studies; mainly in the solvent ILiF-NaF-KI', that corrosion product ions would not interfere with the proposed determination, experimental verification has not been available until this period. 'The molar absorptivity of Fe(Il), Cr(II), and Ni(II) at their wavelengths of maximum absorbance 1s 5 at 1020 nm, & at 760 nm, and 10 at 432 nm respectively. None of these dissolved species will interfere at concentration levels of 10 to 100 times that expected in the fuel salt. In general, the spectra of these 3d jons can be interpreted as arising from essentially octahedral coordination in the case of Ni(II) and Cr(II) and distorted octahedral symmetry in the case of Fe(IT). A cursory spectral study of Cr(III) was made. The molar absorptivity et 1ts wavelength of maximum absorbance, 706 nm, is approxi- mately 7; Cr(III) is not expected to be present in the fuel salt. Tn a study of the ahsorption spectrun of Fe(II) in several different IiF-Bel's soluticns, an extranecus peak was noted at 432 nm. The position of the peak suggested an Ni(IT) jmpurlty in the melt. Based on a subse- quent spectral study of Ni(II) spectra in these melts, a molten~-salt spectrophotometric analysis of Ni{II) concentration was possible. The comparison of this determination to wet chemiceal analysis of the same samples is given below. Ni(II) (% w/w) Molten-Salt Spectra Wet Chemical P il Sample Fe-IBog 0.21 0.19 FE"IBBB 0.18 0-22 195 These melts were made in graphite contalners but were stirred with a nickel stirrer. It would seem that this high nickel contamination, found originally by the mollen-salt spectra, may have arisen from the stirrer. An extremely sensitive absorption peak attributed to U{IV) which causes total absorption of most ultraviolet light has been found at 235 nm. The molar absorptivity of this peak is approximately 1500. The peak has been observed in aqueous abscorption spectra st 207 nm,39 but has not been reported previcusly in any molten-salt solvent. A possible analyt- ical application of this absorption would be a continuocus spectrophoto- metrlic monitoring of coclant salt for leakage of uranium-bearing fuel salt. Voltammetric and Chronopotentiometric Studies of Uranium in Molten ILil- BeF,-ZrF, — D. L. Manning and Gleb Mamantov* Electrochemical reduction and oxidation of U(IV) in IiF-BeF,-ZrF, (64-34-1.8 and 65.6-29.4-5.0 mole %) was investigated by rapid-scan voltammetry and chronopotentiometry. Well-defined and reproducible current-voltage curveg and potential-time curves were cobtained at concen- trations of uranium as high as 0.8 mole % (MSRE fuel). From the data obtained so far, the results are very encouraging from the standpoint of utilizing the voltammetric approach as a means for in-line monitoring of uranium in molten Tluoride systems of interest to the molten-salt reactor program. Utilizing a platinum indicator electrode, the relative standard deviation of peak current Tor 41 runs over a period of 36 days was 2.0%; considerably better precision (standard deviation 0.76% for 8 runs) was ocbtained over a period of 2 hr. Other indicator electrodes tested in- cluded pyrolytic graphite, molyodenum, tungsten, and tantalum. Of the electrodes tested, better reproducibility was obtained at platinum or platinum-10% rhodium. At 500°C the reduction of U(LV) at platinum is a reversible one- electron process, as determined from Nernstian log plots and the diag- nostic criteria of linear sweep voltammetry. Alsc from chronopoten- tiometry, a plot of the current-density—transition-time product (igT) vs (transition time)l/z vielded a straight line, which is in agreement with theory for a reversible electrode process. TFrom the slope of the line, the diffusion coefficient for uranium was calculated to be about 1.5 X 1078 em?/sec, in good agreement with the value obtained by voltammetry (1.5 to 2.0 x 10™° cm?/sec). The effect of temperature on the diffusion coefficient was determined over the range 480 to 600°C; and from a plot of log D vs 1/T the activation energy which corresponds to the reduction of U(IV) to U(III) was found to be approximately 11 kcal/mole. Additional data are being collected to evaluate further the precision and reproducibility of the voltammetric measurements at different levels *Consultant, Department of Chemistry, University of Tennessee, Knoxville. 196 of uranium concentrations, from which it is hoped tec obtain a better assessment of this approach as an analytical method for in-line deter- mination of U(IV) in molten fluorides. A new voltammeter is being bullt that will measure a 20-fold higher current {100 ma) than existing equipment, so that electrodes with more reproducible area can be used. With present equipment the electrode is limited to a 20-gage platinum wire inserted only 5 mm into the fuel, so that slight changes in melt level Introduce a significant change 1n re- duction current. The new instrument will alsc provide a more rapld voltage scan, up to 500 v/sec, to minimize flow effects in an in-line cell. In-Tine Test facility — R. F. Apple, J. M. Dale, J. P. Young, and A. 5. Meyer In view of the value and potential success of methods for the con- tinuous analysis of circulating salt streams, a facility to provide salt streams of known composition is being considered to provide a test of equipnent under similated in~line measurements. The most practical facility commensurate with the available space 1n a California hood con- sists of a 20-kg salt reservoir fitted with a stirrer, ports for sampling and for the addition of solid and molten constituents, purge streams and electrodes for purification, and a helium gas 1ift to continucusly transfer a stream of salt to an elevated constant-level vessel. From the constant- level vessel, salt streams can be gravity-Ted to apparatuses for testing electrometric and spectrophotometric techniques of analysis and for deter- mining parameters for the continuous determination of oxide by counter- current eguilibration of a salt stream with anhydrous hydrogen fluoride. The analyzed salt stireams will then be returned to the reservoir. This facility will also be used to test capillary and orifice techniques Tor the control of low salt flows and to design small freeze valves. Tests of various gas 1ift designs are being carried out with simulated molten fuel (aquecus zinc chloride solutions) to determine whether a 6- to §-ft 1ift is practicsal. Analysis of Helium Blanket Gas — C. M. Boyd, C. A. Horton, A. D. Horton, and A. S. Meyer The Analytical Chemistry Division, together with members of the Re- actor and Reactor Chemistry Divisions, participated in various experi- ments to determine possible sources of the organic depesits which have caused pluggling of valves and filters in the MSRE off-gas system. The analytical support included: 1. +the installation of a continuous hydrocarbon analyzer to monlitor gas streams from Reactor Chemistry experiments4o to simulate oil leaks into the MSRE pump and to evaluate trapping systems for reducing hydrocarbon concentration, and to monitor hydrocarbons in the off- gas from the Y-12 pump test 1oop;41 197 2. the determination by gasg chromatographic analysis and by selective chemical reactions of individual hydrocarbons in "grab" samples from the above experiments and in a trapped sample from a nonradio- active MSRE purge line;42 3. the development of a technique to measure the concentration of hydro- carbong collected on an experimental charcoal trap by pyrolysis of a sample of the charcoal followed by gas chromatographic analysis of the pyrolyzate. By sampling the charcoal bed at various depths, it is possible to determine the distribution of individual hydrocarbons as a function of trap length. The continuous hydrocarbon monitor proved to be the most ussful of these technlgques, particularly in measurements performed at the Y-12 test loop with P. G. Smith, of the Reactor Division, and R. G. Ross, of the Reactor Chemistry Division. Figure 7.23 is a flow schematic of the in- Jection experiment. Although the experiment was designed to measure the effects of deliberate injection of oil into the pump tank, the initial results revealed an actual oll leak in the pump. The experiment there- fore afforded an opportunity to study the oil leak and correlate the hydrocarbon level in the pump tank off-gas with an operating variable (Ap across the shaft annulus) and thus distinguish between possible leak locations. A typlcal plot of the hydrocarbon concentration in the pump tank off-gas is shown in Fig. 7.24. When the pump rotation was stopped, the shaift annulus Ap decreased from 3 to 0.5 psi, and the hydrocarbon level in the off-gas dropped immediately to less than 10 ppm methane egquivalent. After about 25 hr the Ap began to recover, and the hydro- carbon level started to undergo & series of rapid excursions that are suggestive of discrete drops of oil entering the hot regions of the pump tank. The average hydrocarbon concentration gradually returned to a level ORNL--DWG, 66-6827A SAMPLE POINT 1 He - LN 41L 2 3 SAMPLE POINTS PUMP TANK PURGE A e ORNL-DWG &5-10029A 1200 — —$}: I e T T T T : e ,_, ~ : : o = 1000 +~ — — - - - o '_ g | S 2 800 — iy = © . oz : S 3 i 2 Lo i R 2 600 ‘ i xr o < £ O @ 2 € ao0 o € - o T & ; | I o ;I’ 200 p— — f+— o = ]_ O ! = ; P 0 — —— Do T o o ] 210 220 230 240 260 270 280 TEST TIME (hr! Tig., 7.24. Hydrocarbong in Y-12 Pump Tank Off-Gas. of about, 300 ppm, in parallel with inecreasing Ap. Similar fluctuations in hydrocarbon concentrations were cbserved during three additional changes in pump operation to reduce the Ap, which was ultimately shown to be caused by a salt plug at the bottom of the shaft annulus. Because this differential pressure is exerted across the seals between the shield plug and the catch basin, the position of the leak was defined as through these seals rather than down the shaft annulus. This was confirmed, on disassembly, by an oil film and pyrolysis stain on the outside of the shield plug. Injection of oil, Gulf Spin 35, into the pump tank showed that es- sentially all the oil entering the tank appeared as hydrocarvons 1n the off-gas. Table 7.18 shows gas chromatographic analyses of the off-gas together with those from helium effluents from Reactor Chemistry experi- ments in which the oil was injected into a helium stream entering an empty nickel pot at 600°C. These results revealed that, at least in the absence of radioactivity, oil entering the pump tank is predominantly cracked to light hydrocarbons, methane, ethane, and unsaturates lighter than Cs5, which are trapped ineffectively on charcoal traps at 100°C. Conversely, in the Reactor Chemlstry experiments the cracking was incom- plete to yield a substantial fraction of > Cg hydrocarbons, which are trapped with high efficiency. A thermal conductivity method to measure the total hydrocarbon con- centration in the radiocactive off-gas of the MSRE continuously has been developed with the Reactor Chemistry Division. In this method the off- gas sample is passed over copper oxide at 700°C to convert hydrocarbong to COs and H,O. This oxidized stream is passed through one side of a thermal conductivity detector and thence to a trap containing Ascarite and Mg(C10z), to yield a stream of inert gases which is directed through the reference side of the thermal conductivity cell. In bench-top tests the response of this apparatus was linear to total hydrocarbon concentra- tion of 1000 ppm with a limit of detection below 10 ppm. A similar sys- tem with the trap and copper oxide furnace designed for one year of reac- tor operation will be installed in the gas-sampling station of the MSRE. 199 Table 7.18. Hydrocarbons Produced by 0il Injection Injection rate, 16 cm®/day Hydrocarbon Concentration {(ppm methane equivalents) Y-12 Test Loop Hy drocarbon Reactor Chemlstry Simulated Ieak 2 hr After 20 hr After Tests Start of Injection Start of Injection Before After Before After Before After Trap Trap Trap Trap Trap Trap CHy, 25 30 210 210 320 250 CoHg &4 12 tidy 42 46 40 CoH, 140 140 600 590 700 690 C3Hg 100 120 170 158 230 210 Unsatd. Cu 56 94 98 3 130 160 Unsatd. Cs 78 4 10 7 24, A and Cg Aromatics 158 206 > Cg 300 Total 703 400 1290 1010 1666 1354 7.6 Development and Fvaluation of Equipment and Procedures for Analyzing Radioactive MSRE Salt Samples F. ¥X. Heacker C. . Tamb L. 7. Corbin The remote apparatus for determining the oxide content of MSRE salt samples was installed In cell 2 of the High-Radiation-ILevel Analytical Iaboratory (ARLIAL) (Building 2026). Fluoride salt samples and tin oxide standards were analyzed to check the apparatus and Tamiliarize the HRIAL persomnel with the method. The results were satisfactory, and the phys- ical manipulations required were performed adequately. DBight 50-g fuel- salt samples have been analyzed remotely. The time required to decontaminate the transport containsrs was greatly reduced py using disposable mild steel plugs for each sample submitted. 200 Sample Analyses From January 1, 196G, through June 30, 1966, 43 MSRE fuel-salt sam- ples were submitted for analysis. The samples were analyzed as shown below. Analysis Number of Determinations Uranium 35 Zirconium 35 Chromium 35 Beryllium 35 Fluorine 35 Iron 35 Nickel 35 Molybdenum 7 Lithium 35 RCA Prep. 22 MSA Prep. 5 Oxide & Carbon 3 Of the 43 samples submitted, two 50-g oxide samples were not analyzed due to contaminated ladles. Tlhree samples were analyzed for carbon and found to contain <50 ppm. Several of the samples were received with a silver and a Hastelloy N wire colled onte the stainless steel cable between the latch and ladle. The latch, wires, and cable were separated and prepared for radiochemical analyses. Quality-Control Program The quality-control program was continued during the first and second guarters of 1966. A composite of the values obtained by four different groups of shift personnel is shown in Tables 7.19 and 7.20. Molybdenum values are not shown since it was not added to the synthetic solutions. It was evident from the second-guarter control data that a positive hbias of approximately 3% existed in the amperometric zirconium methaod. The bias was attributed to poelymerization of the zirconium in the standard solutions used to standardize the cupferron titrant. Although the hias appears to have been eliminated by the preparation of new zirconium standards, more data are needed o confirm it. The positive bilas existing in the spectrophotometric nickel method has been rediced to approximately 8% by cnanging the color filter in the filter pnotometer and constructing a new calibration curve. Further ef- forts are being made to e¢liminate the bias completely. 201 Table 7.19. BSummary of Control Results, January Through February 1966 Limit of Error (%) Number of Determinations Determination Method Determinations Fixed Found Beryllium Fhotoneutron 27 5.0 2.43 Chromium Amperometric 25 15.0 7.95 Iron o-Phenanthroline 34 15.0 7.93 Nickel Dimethylglyoxime 30 15.0 12.74 Uranium Coulometric g 1.0 1.16 (high sen.) Zirconium Amperometriec 30 5.0 5.74 Table 7.20. Summary of Control Results, April Through June 1966 Limit of Error a Determination Method Number of (%) Fixed Found Beryllium Photoneutron 26 5.0 2 .60 Chromium Amperometric 36 15.0 12.46 Iron o-Fhenanthroline 41 15.0 7.5% Nickel Dimethylglyoxime 48 15.0 7.67 Uranium Coulometric 127 1.0 1.04 (high sen.) Zirconium Amperometric 57 5.0 5.48 n 10. 11. 12. 14 15. 17. 18. 19. 20. 21. 22 . 23. 24 25. MSR Program Semiann. Progr. Repb. Feb. 28, 1966, ORNL-3936, p. 122. Tone operational history of these runs is described in sect. 1.1 of this report. 5. 5. Kirslis, this report, sect. 7.4, subsection entitled "Fission Product Behavior in the MSRE." R. 7. Apple, J. M. Dale, and A. S. Meyer, this report, sect. 7.5, subsection entitled "Determination of Oxide in Radiocactive MSRE Samples." R. E. Eby, ORNL Isotopes Division. R. E. Thoma to P. N. Haubenreich, "Chemical Analysis of MSRE Flush and Ccolant Salts in Prenuclear Test Period," MSR-65-19 (March 19, 1965) (internal correspondence). 5. Cantor and W. T. Ward, Reactor Chem. Div. Ann. Progr. Rept. Dec. 31, 1965, ORNL-3913, pp. 27-28. L. Shartsis and 5. Spinner, J. Res. Natl. Bur. Std. 46, 176 (1951). J. D. MacKenszie, J. Chem. Phys. 32, 1150 (1960). J. H. DeBoer and J. A. M. Van lLiempt, Rec. Trav. Chim. 46, 124 (1927). L. J. Kinkenberg, Rec. Trav. Chim. 56, 36 (1937). Structure Reports, vol. XITI, p. 342. J. D. Edwards et al., J. Electrochem. Soc. 100, 508 (1953). S. Centor, Reactor Chem. Div. Ann. Progr. Repb. Dec. 31, 1965, ORNL-3913, pp. 29-32. S. Cantor and W. T. Ward, ipid., p. 27. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNI~3936, pp. 141—45., D. G. Hill to W. R. Grimes, private communication, June 29, 1966. D. G. Hill tc W. R. Grimes, private communication, July 1, 1866. W. 5. Ginell, Nucl. Tech. 51, 185 (1959). J. J. Egan and R. N. Wiswall, Nucleonics 15, 104 (1957). Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1964, ORNL-3591, p. 50. Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, p. 16. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, p. 145, Zr0p was prepared by H. H. Stone, Reactor Chemistry Division. MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-39356, p. 147. 26 . 27, 28. 29. 30. 31. 32. 37. 38. 39. 40, 41, 42 203 R. P. Elliott, Constitution of Binary Alloys, First Supplement, p. 203, MeGraw-Hill, New York, 1965. MSR Program Semiamn. Progr. Rept. Feb. 28, 1966, ORNI~3936, p. 148. W. D. Burch, G. M. Watson, and H. 0. Weeren, Xenon Control in Fluid Fueled Resctors, ORNL-CF-60-2-2 (July 6, 1960). I. Spiewak, "Xenon Transport in MSRE Graphite," M3R-60-28 (Nov. 2, 1960) (internal correspondence). G. M. Watson and R. B. Evans III, Xenon Diffusion in Graphite; Ef- fects of Xenon Absorption in Molten Salt Reactors Containing Graph- ite, ORNL-CF-61-2-59 (Feb. 15, 1961). J. W. Miller, Xenon Poiscning in Molten 5alt Reactors, ORNI~CF-61- 5-62 (May 3, 1961). R. B. Evans III, "Xenon Poisoning in Molten Salt Reactors Contain- ing Graphite," Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1961, ORNI~3127, pp. 15-16. P. R. Xasten, E. 3. Bettis, and R. C. Robertson, Design Studies of 1000-Mw(e) Molten-8alt Breeder Reactors, ORNI-3926 (August 1966). B. Manowitz, quoted by P. G. Salgado in Nuclear Reactor Magneto- Hydrodynamics Power Ceneration, LA-3368, p. 23. MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNI-3872, p. 127. A. M. Perry, internal memorandum, July 30, 1966. MSR_ Program Semiann. Progr. Rept. Fep. 28, 1966, ORNI~3936, p. 154. MSR Program Semiann. Progr. Repb. Aug. 31, 1965, ORNL-3872, p. 145. Donald Cohen and W. T. Carnall, J. Phys. Chem. &4, 1933 (1960). B. F. Hitch, R. G. Ross, and H. F. MeDuffie, Tests of Various Parti- cle Filters for Removal of 0il Mists and Hydrocarbon Vapors, ORNI~ T™-1623 (to be issued). ' A. 5. Meyer, Investigation of Hydrocarbons in the 0ff-Gas from the Y-12 Test Loop, ORNI~CF-66-8-30 (September 1966). A. 5. Meyer, "Hydrocarbons ian Seal Purge of MSRE Pumps," MSR-66-10 (May 1966) (interunal correspondence). Part 3. MOLTEN-SALT BREEDER REACTOR STUDIES 8. MOLTEN-SALT BREEDER REACTOR DESIGN STUDIES Paul R. Kasten E. 8. Bettis Roy C. Rcbertson J. H. Westsik H. F. Bauman H. T. Kerr The MSBR reference design concept was presented previously,l and is a two-region, two-fluld system, with fuel salt separated from the blanket salt by graphite tubes. On-site fuel recycle is employed, using fluoride-volatility and vacuum-distillation processing., In additiocn, results were given for the case of direct removal of protactinium from the blanket region; the associated concept was termed MSBR(Pa). Since these studies, additional design conditions were investigated and evaluated,? and these are discussed below. 8.1 Design Changes in MSBR Plant In reactor design studies it often occurs that certain features of the detaliled design undergo changes as more understanding is obtained of the overall problems and as new ways are discovered to solve a given de- sign problem, §Such changes have tsken place during the M3BR design studies; the most important are those associated with the primary heat exchanger designs and the pressures that exist in the wvarious circulating- salt systems. An objectionable feature of the MSBR heat exchanger design considered previously was the use of expansion bellcows at the bottom of the exchanger. These bellows permit tubes In the central portion of the exchanger to change in length relative to those in the annular region due to thermal conditions. Since such bellows may be impractical to use under reactor operating conditions, a new design was developed that eliminated them. Figure 8.1 shows the revised heat exchanger design. The expansion bellows were eliminated, and changes in the tube lengths due to thermal conditions are accommodated by the use of sine-wave type of construction, which permits each tube to adjust to thermal changes., In addition, the coolant salt now enters the heat exchanger through an annular volute ab the top and passes downward through a baffled outer annular region. The coolant salt then passes upward through a baffled inner amnular region and exits through a central pipe. In Mig. 8.1 the flow of fuel =slt through the pump is reversed from that given previcusly® in order to reduce the pressure in the graphite Tuel tubes., TFuel salt enters the heat exchanger in the inner annular region, passes downward through the tubes, and then flows upward through the tubes in the outer annular region before entering the reactor. The blanket-salt heal exchanger was also revised to give 2 design similar to that of Fig. 8.1. The general features of these exchangers and their placement in the reactor cell are shown in Fig. 8.2. The 207 208 ORNT, DWG. 66-713% —FUEL LEVEL (DUNP) FUEL SALT DUMP TANK I o —— — —— FUEL SALT PUMP FUEL SALT 1O k b T I ‘ & FROM REACTOR |+ | i N N COCLANT SALT FROM ! ,,;-7“-' { STEAM HEAT EXCHANGERS rd / ; ’j:u-:.-xfl | e \\ W i ‘I —COOLANT PASS | ] 1l | ALY SEPARATING BAFFLE | ; ‘ it A 1f | | ok ‘ [4-4 i ;-cl u_f b Wn T e lflnjt i B L i N 1 I g AL FLOW ARRANGEMENT I il FUEL SALT-IN TUBE SIDE fo DU L ‘ L ‘L ML COOLANT SALT-IN SHELL SIDE ] 1 T b | ‘ T i \ | i 4% : \ ‘ “ e —';'ll’lol.—' ; i | il | ’ | L)y aH L8 0" DD il i R TRl | ! B \ | | | | Il 1t =i = \‘ { 7.):#?: : 't“ i ; D ! ‘ | : | ‘ ' : | \ . I L0 ! VH’U’“” 1 Jl ‘{H il ] f\ Rlefeld 1 f\ii;j \\_ - /f | ”\\.t___; _,7/. 10 FUEL———— i DRAIN TANKS T \ ~COOLANT SALT TO BLANKET HEAT EXCHANGER Fig, 8.1. Revised Fuel Heal Exchanger for MSBER. 209 ORNL DWG. 66-7109 FUEL PUMP BLANKET PUMP MOTOR CONTROL ROD MOTOR DRIVE CONSTANT - SUPPORT HANGERS FUEL DUMP TANK WITH COOLING et = COILS FOR _.'/.) o AFTER HEZAT S REMOVAL s o ABLANKET 72 S THEAT e JEXCH. = YL Smee———m——— - s el e . [ ' 4 C o > & -5 s, - 2 > 4 ______ a1 s 3 ¥ a a o . 2 ~ AR < e — e « |CORE .;;u REACTOR - VESSEL i - , nc i e 2 i L. FUEL SALT : IDIST. : & oo+ PLENUMS e € & L - - . . S -~ . e e . ‘,,‘ . ‘ PRIMARY = HEAT - EXCH. e g :1.‘ 1 ‘o .S ] R . w Ny TR AT B B g T Y L LREACTOR CELL HEATERS h, . R R AR L et ‘ -57i0 h e 020 # [ g 750.5224 STEAM COOLANT SALT Lbs/ hr Los/in® aos. Btu/ib °F NOTE CONDITIONS FOR "8" T e I TN S IS SEDED S SE— — DTSR i 433~ 4512n EQUIPMENT (EXCEPT TUR-GEN) SAME AS FOR ‘A", Fig. 8.0. 5055,950# A L] . -~ — 540P-1000 —-i518 5h 3515 F' IOOO | ! ! | | } | | | | 1314.0 h 1,579,000 # |46P - 6937 4 I4240h | | 1P TURBINE GENERATOR "A" GROSS OUTPUT r__ 530.4 Mwe T ORNL, DWG. 66-7134 CONDENSER A" | 1.5 " Hg abs 32,660 # 103.0%-710h HOTWELL PUMPS )IA‘I\ | CONDEMSATE 4 ! BOOSTER | | PLMPS - —_——— e | —— e 2O !_ T - ; 1000 3h i i l | T:;—Mfl GENERATOR "8" i GROSS [ OUTPUT f 505.C Mwe ! '/r—’/r\l\h*] | | | Pyt | ; ' p | '*i Py e Lt Lt ..E_'.LJI — L% — :J__'"i e _d T snzos 405,330 # [362P-aga e 1242 7h 1,560,480 # 220, 460% 321,8504 SUMMARY OF PERFORMANCE AT NET 146.6 ! g i i L 32B.5-12040h T T : - T T — | 1 I 4-3 _i I 6-8 ! 1 | | q8898004 | [ 1 | 2 ~— f 77 l I I : DEAERATOR | | | i b ! g ! | : i ; l n ;;J‘:_ _‘fi STORAGE | L_i . L_.____._‘ 953,190+ ' ’ | | 3761-349.3 1 = | | @;’i‘l";s. | (357553295 2341° M 1734143 h | | 202.4n | ] |— 7es0s504 | | 52202304 | 141,900% | 183.4°- (5.3 h QUTPUT -1H85h 3BOOP -366.1°- 343.3h Alternate Steam System Flowsheet for 580°F Feedwater, ELEC. POWER FOR AUX BF PRESSURE -BOOQOSTER PUMPS GROSE GEN. QUTPUT 8F PUMPS REACTOR HEAT INPUT GROSS E£FF OF CYCLE NET EFF OF CYCLE NET HEAT RAYTE 1436 -i11.5 h RATED LOAD 10097 iAwe 25.7 Mwe NONE 1035. 4 Mwe 30.6 Mw 2225 Nwt 47.9 % 454 % 7,518 Btu/Kwh 219 Table 8,5. MSBR Steam System Design and Performance Data for Case A and Case B Conditions Cagse A - MSBR Steam Cycle Case B — MSBR Alternative Total steam capacity, lb/hr Temperature of inlet feed- water, °F Enthalpy of inlet feedwater, tu/1b Pressure of inlet feedwater, peia Temperature of exit gteam, °F Pressure of exit steam, psis Enthalpy of exit steam, Btu/1b Temperature of inlet coolant salt, °F Temperature of exit coclant salt, °F Average specific heat of coolant salt, Btu 1b™1 °p™1 Total coolant-salt flow 1b/hr cfs gpm 10.068 % 10° 700 ~3800 1003 ~3600 1424.0 1125 850 0.41 58,468 x 10° 129.93 58,316 with 700°F Steam Cycle with Feedwater 580°F Feedwater General performance Reactor heat input, Mw 2225 2225 Net electrical output, Mw 1000 1.009.7 Gross electrical generation, 1034.9 1035.4 Mw Station auxiliary load, Mw 25.7 25.7 (electrical) Boiler~feedwater pressure- 9.2 None booster pump load, Mw (electrical) Boiler~feedwater pumnp steam- 29,3 30.6 turbine power output, Mw Flow to turbine throttle, 7.152 x 10° 7.460 % 10° 1b/hr Flow from superheater, 1b/hr 10.068 x 106 7.460 x 108 Gross efficiency, % 47.83 47,91 Cross heat rabe, Btu/kwhr 7136 7124 Net efficiency, % 4,9 45,4 Net heat rate, Btu/kwhr 7601 7518 Beiler-superheaters Number of units 16 16 Total duty, Mw (thermal) 1931.5 1837.0 7,460 % 10° 580 583.6 ~3800 1003 ~3600 1424, 0 1125 850 Q.41 55.608 x 10° 123,57 55,463 Table &8.5. (continued) Case A — MSBR Steam Cycle with 700°F Feedwabter Cagse B ~ MSRR Alternative Steam Cycle with 580°F TFeedwater Steam reheaters Number of units Total duby, Mw (thermal) Total stesm capacity, 1b/hr Temperature of inlet steam, °F Pressure of inlet steam, psi Fnthalpy of inlet steam, Btu/1b Temperature of exit steam, °F Pressure of exit steam, psia Enthalpy of exit steamn, Btu/1b Temperature of inlet coolant salt, °F Temperature of exit coolant salt, °F Average specific heat of coolant salt, Btu 1b™1 °F™ Total coolant-salt low 1b/hr cf's o Coolant-calt pressure drop, inlet to ocutlet, psi Reheat-steam prenheater Nuntber of units Total duty, Mvw (thermal) Total heated steam capacity, 1b/hr Tnlet temperature of heated steam, °F Exit temperature of heated steam, °F Tnlet pressure of heated stean, psia Exit pressure of heated steam, psia Inlet enthalpy of heated steam, Btu/lb Exit enthalpy of heated steam, Btu/1b Total heating steam, 1b/hr Tnlet temperature of heating steam, °F 8 293.5 5,134 x 10° 650 ~570 1323.5 1000 ~54.0 1518.5 1125 850 0.4 8.884 x 10° 19,742 8861 ~50 8 100.45 5.134 x 10° 551.7 650 ~580 ~570 1256.7 1323.5 2.915 x 10° 1000 g 388.0 5.056 x 10° 551.,7 ~500 1256.7 1000 ~54.0 1518.5 1125 850 0.41 11,744 % 108 26.098 1714 ~60 None 221 Table 8.5. (continued) Case A - Case B — MSEBR Steam Cycle MSBR Alternative with 700°F Steam Cycle with Feedwater 580°F Feedwater Exit temperature of heating 866 steam, °F Inlet pressure of heating 3515 steam, psia Exit pressure of heating steam, psia Boller-feedwater pumps Nurmber of units 2 2 Centrifugal pumps Number of stages Centrifugal pump Feedwater flow rate, 1b/hr total 6 6 10,067 x 108 Feedwater flow rate, 7152 x 10° 7460 % 10° 1b/ar total Required capacibty, gpm 8060 8408 Head, f% ~9380 ~2380 Speed, rpm 5000 5000 Water inlet temperature, °F 357.5 357.5 Water inlet enthalpy, Btu/1b 329.5 329.5 Water inlet specific volume, ~Q, 01.808 ~0,01808 ££2 /1o Steam~-turbine drive Power required at rated 14,66 15.30 flow, Mw (each) Power, nominal hp (each) 20,000 20,000 Throttle steam conditions, 1070/700 1070/700 psiaf°F Throttle flow, 1b/hr (each) 413,610 431,400 Exhaust pressure, psia ~{7 ~ 77 Number of stages 8 & Number of extraction points 3 3 Boiler-Teedwater pregsure- booster pump Number of units 2 None Required capacity, gom {(each) 2500 Head, ft ~1413 Water inlet temperature, °F 695 Water inlet pressure, psia ~3500 Water inlet specific volume, ~(. 03020 £t2/1b Vater outlet temperature, °F ~700 Electric-motor drive Pover required at rate flow, 4. 587 Mw (each) Power, nominal np (each) 6150 222 The elimination of the feedwater pressure-booster pumps required in case A saves 2bout 9.2 Mw (electrical) of auxiliary power, which, to- gether with the improvement in the cycle thermal efficiency due to the additional stage of feedwater regeneration, makes about 9.7 Mw (electrical) additional power available from the case B cycle., The overall net thermal efficiency is thus improved frowm the 44.9% obtained from case A to 45.4% in case B, To complete the discussion of case A vs case B conditions, the cost estimates for the affected items of equipment were compared; the results are summarized in Table 8.6. As shown, the case B arrangement reguires about $465,000 less capital expenditure, primarily due to removal of the pressure-booster pumps. [In this cost study 1t was assumed that the 580°F liquidus-temperabure coolant salt has the same cost (about $1.00/1b) as the MSBR coolant salt.] The lower construction cost reduces power costs by about 0.008 mill/kvwhr (electrical), while the increased efficiency lowers power cost by about 0.026 mill/kwhr (electrical) (private [inanc- ing), to give a total saving of about 0.034 mill/kwhr (electrical) [0,021 Table 8.6. Cost Comparison of 700°F and 580°F Feedwater Cycles for MSBR Number of Case A — 700°F Cagse B — 580°F Units Feedwater feedwater Feedwater pressure-booster 2 $ 400, 000 None pumps Reheat-steam preheaters S) 180,000 Neone Special mixing tee B 5,000 None Feedwater heater No, O None $ 150,000 Charge for extra extraction None 45,000 nozzle on turbine for heater No. 0O Boiler-superheaters 16 6,000, 000° 5,900,000 Reheaters g 2,720,000" 2,880,000 39,305,000 £8,975,000" Cost differential Direct construction cost $320, 000 Total construction cost $465,000 a, , - . ; Table shows only those costs different in the two cycle arrangements and is not a complete listing of the turbine plant costs. - 0 . . o The high-pressure feedwater heater added in case B was designated " "n = . 0 , . . No. 0" in order not to disturb the heater numbers used in case A. CEstimated on basis of $130/ft2. dEstimated on basis of $140/ft2. “Gstimated on basis of $125/7t2, T . Tndireect costs were assumed to be 41% of The direcet costs. 223 mill/kwhr (electrical) for public financing]. This saving in a 1.000- Mw (electrical) plant (0.8 load factor) corresponds to aboub $238,000 per year. The present worth (6% discount factor) of this saving over a 25-year period is about $1.5 million, For several MSBR power plants, the saving would be proportionally greater. Thus, there is an economic incentive Tor developing a coolant salt with a low liguidus temperature, zo long as its inventory cost does not outweigh the potential saving. Tf the inventory cost of the coclant salt for case B were about $2.4 million more than +that for case A, the potentisl saving would be canceled by the increased coolant-salt inventory cost (for a privately owned plant). 8.4 Additional Design Concepts Other molten-salt reactor designs were studied briefly.? In general the technology required for these alternative designs is relatively un- developed, although there are experimental data that support the feasibility of each concept. An exception is the molten-salt converter reactor (des- ignated MSCR), whose application essentially requires only scaleup of MSRE and associated Tuel-processing technology. However, the MSCR iz not a breeder, although it approaches break-even breeder operation. The ad- ditional concepts are termed MSBR({Pa-Pb), SSCB(Pa), MOSEL{Ps~Fb), and MSCR. The MSBR(Pa-Pb) designation refers to the MSBR{Pa) modified by use of direct~contact cooling of the molten-salt fuel with molten lead. Lead is dmmiscible with molbten salt and can be used as a heat exchange medium within the reactor vessel to significantly lower the Tisslle inventory external to the reactor. The lead also serves as a heat transport medium between the reactor and the steam generators. The SSCB(Pa) deszignation refers to a Single-Stream-Core Breeder with direct protactinium removal from the fuel stream.” This is essentially a single-region reactor having fissile and fertile material in the fuel stream, with protactinium removal Trom this stream; in addition, the core reglon is encloged within a thin metal membrane and is surrounded by a blanket of thorium-containing salt. Nearly all the breeding takes place in the large core, and the blanket "catches" only the relatively small fraction of neutrons that "leak" from the core (this concept iz also referred to as the one-and-one-half region reactor). The MOSEL(Pa~Pb) designation refers to a MOlten-Salt Epithermal, breeder having an intermediate-to-fast energy spectrum, with direct protactinium removal from the fuel stream and direct-contact cooling of the fuel region by molten lead. No graphite is present in the core of this reactor. The M3CR refers to a Molten-Salt Converter Reactor that has the Tfertile and fissile material in a zingle stream. No blanket region is o employed, although a graphite reflector surrounds the large core. The fuel-cyele performance characteristics for these reactors are summarized in Table 8.73; in all cases the methods, analysis procedures, Table 8.7. Summary of Design Conditions and Fuel-Cycie Performance for Reactor Designs Studied Design Concitions . . . & Reactor Designation MSBR(Pa) MSBR MMSEBR(Pa.) MSBR(Pa-Pb) SSCB{Pa) MOSEL(Pa-Fb) MSCR Dimensions, ft Core h C Height 12.5 12.5 7.9b 12.5 16,0 3.OC 20.8 Diameter 10,0 10.0 6,3 10.0 9.8 6.5 15.5 Bianke®t thickness Racdial 1.5 1.5 2.0 i.5 1.2 3.0 Axial 2.0 2.0 2.0 2.0 0.0 Volumne fractions, core Fuel J.169 0,169 0.17 0.16%9 0.185 0.5 0.105 Fertile 0.073 0.074 0.05 0.076 0.0 0.0 0.0 Rfi Moderator 0.758 0.757 0,78 0.755 0.807 0.0 0.895 A Salt volumes, ft° Fuel Corse 166 166 166 166 230 63.5 4’76 External 551 547 574 120 60C 0.7 654 Total 717 713 740 276 830 E4.2 113C Fertile, total 1217 3383 1570 1324 G835 758 0,0 Fuel-salt composition, mole % LiF 63.6 63.6 63.6 63.6 71.0 7L.0 70.0 Beiy 36.2 36.2 36.2 36.2 23,1 0.0 13.0 Th¥, 0.0 0.0 0.0 J.C 8.68 24.0 26.55 UF,, {fissile) 0.22 0.23 ¢.21 0,23 0.23 5.0 0.45 Core atom ratios Th/U 41,7 39.7 28.4 41,5 37.7 4,76 36.7 C/U 5800 5440 5980 5520 6280 0.0 6525 Teble 8.7. (continued) . . . a Reactor Tesignaticn MSTR(Pa) MSBR MMSER(Pa} MSBR{Pa-Ph) SSCR(Pa) MOSEL(Pa-Pb) MECR Power density, core averags, ww/liter Gross 80 80 80 80 66 618 17 Tn fuel s< Ak 475 473 A2 341 1236 165 Heytron flux, cor newtrons o2 s Thermal 7.2 x 104 6.7 X 10%% 7.3 x 104 6.8 x 1024 6.1 X 10%% 0.0 » 10%4 1.9 x 1034 Pasth 12,1 X 10%% 12,1 x 104 11,7 % 10%% 12,1 x 104 10,0 x 1014 72.2 % 1014 2.7 X 104 Fast, over 100 kev 3.1 x 1034 3.1 x 104 3.0 x 1034 3.1 % 104 2.6 X 1014 23.3 x 1014 0.7 % 1014 Neubron preduction per 2.227 2,221 2.229 24220 2.226 2.260 2,201 ahsoroticn {ne) Tuciesr and fuel-cycle perfcrmance Fuel yield, %/year 7.95 4,86 7.21 17.2 6,63 16.3 Breeding ratio 1.07 10.5 1.07 1.08 1.06, 114 0.9 N Fuel-cycle cost, mills/kvhr 0.35 0.46 0.38 0.25 0,37 0.13 0.57 W Specific fiseile inventory, 0.68 Q.77 G.76 0. 34 0.68 0.99 1.63 xg/Ma (electrical) = b “See text for explanation of resctor designations. The core dlmensionsg for thnis case refer to one module of & four-module station. C . . . s < . o For this case, the core had sanular geometry; the fuel ammulus inside diameter was 3 Tt, and the outside diameter was 6.5 ft. Cn Use of of di inventory to abouw irect-contact lead cooling would lower the iuel-cyele cost to zbout 0.32 mill/kwhr (electrical) and the speeific fissgile 0.41 kg/Mw {electrical). 220 and economic conditions employed were analogous to those used in obtain- ing the reference MSBR design data. In genecral, fuel recycling was based on fluoride-volatility and vacuum-distillation processing; direct prot- actinium removal from the reactor system was also considered in specified cases. The results indicate the potential performance of fluoride-salt systems utilizing a direct-contact coolant such as molten lead and the versatility of molten salts as reactor fuels. They also illustrate that single-region reactors based on MSRE technology have good performance charscteristics. Since the capital, operating, and maintenance costs of the M3CR should be comparable with those of the MSBR, the power-produc- tion cost of an investor-owned MSCR plant should be about 2.9 mills/ kwhr (electrical), based on a load factor of 0.8. However, the lower power costs of the MSBR(Pz) and MSBR plants and their superior nuclear and fuel-conservalbion characteristics make development of tThe breeder reactors preferable. References 1., MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, p. 172. P. R. Kasten, E. 3. Bettis, and R. C. Robertson, Desgign Studies of 1.000-Mw(e) Molten-Salt Breeder Reactors, ORNL-3996 (August 1966). 3. P. R. Kasten, Safety Program for Molten-Salt Breeder Reactors, un- published internal report (July 29, 1966). 9, MOLTEN-SALT REACTOR PROCESSING STUDIES M. B. Whatley A close-coupled facility for processing the fuel and fertile streams of a molten-salt breeder reactor (MSBR) will ve an integral part of the reactor system. BStudies are in progress for obtaining data relevant to the engineering design of such a processing facility. The processing plant will operate on a side stream withdrawn from the fuel stream, which circulates through the reactor core and the primary heat exchanger. For a 1000-Mr (electbrical) MSBR approximately 14 £t of salt will be proc- essed per day, which will result in a fuel-salt cycle time of approxi- mately 40 days.l The probable method for fuel-stream and fertile-stream processing is shown in Fig. 92.1. The salt will Tirst be contacted with ¥y for re- moval of U as volatlile UFg. Purified UFg will be obtained from the filu- orinator off-gas (consisting of UFg, excess Fp, and volatile fission product fluorides) by use of NaF sorption. It may be necessary to dis- card as much as 5% of the salt leaving the fluorinator for removal of fission products such as Zr, Rb, and Cs. A semicontinuous vacuum dis- tillation will then be carried out on the remaining salt for the removal CRNL-DWG 85-1BOIRZA Fo RECYCLE mmrmsrmnnnnscmeececceg - p—) e —- L RECYCLE £ e T MgF, COLD SORBER TRAP FiiTis 3 MaFs 81 ft Yday colD TRAP BLANKET SALY SPENT NGF + MgF ] NaF + Magi, UFg Fs Fa LiF - BeF,~ UF, r 15 $t%/day WASTE . FERTILE SALT I.IF - BeF,—UF, LiF MAKEUP UFg 015 f1¥doy 1 = LiF+Bef, PRODUC T = 221 Vg/duy ?fl_@‘fifi&’!’ RECOMBINER WASTE = 0,059 #¥/day TLi=078 kg day FP's - Fig. 2.1. MSBER Fuel and Fertile Stream Processing. 227 228 of the rare eartns, Ba, Sr, and Y. These fission products will he re- moved from the still ia a salt volume equivalent to 0.5% of the strean. The barren salt, the purified UFg, and the makeup salt will then be re- combined. This step involves reduction of UFg to UFs, mixing of these streams, and sparging the resultant material with an Hp-HF stiream, Fi- nally, the salt mixture may be filtered before return to the reactor. 9.1 Semicontinuous Distillation J. R. Hightower L. BE. MclNeese New measurements of the relative volatilities of NdF; and LakFj in LiF have been made using a recirculating equilibrium still. These values are lower than earlier data by a factor of about 50 and are in a range (around 0.0007) where the proposed distillation step in the MSBR proc- esging plant should work very well. The present concept of the distillation step in the MSBRE processing plant uses a continuocus feed stream and vapor removal for separating rarve-earth Tission products (FP“S) trom the fuel salt; the less volatile rare-~cartn FP's will accumulate in the still pot and will be discharged periodically.’ A measure of the decontamination achieved in this step is the relative volatility of the less volatile FP's compared to the carrier salt. The relative volatility of component A compared to com- ponent B is defined as Y.p = %’;’% ’ ) B where thB = relative volatility of A compared to B, Y = vapor-phase mole fraction, X =z liquid~-phase mole fraction. For systems in which the concentration of compouent A is swall and X is nearly unity, the relative volatility can be approximated by 0 o &Y, /X, (2) To achieve good decontamination from tne less volatile fission products their relative volatilitles must be small. oince the major constituent of the still pot will be LiF, experi- mental measurements have been made with mixtures of rare~earth fluorides in LiF. A cold~finger technigue gave approximate values for the relative volatilities of six rare-earth fluorides with respect to LiF (975 to 1075°C) ranging from 0.0l to 0.05.2 These were high enough to sericusly 1limit the effectiveness of the proposed simple distillation scheme. More accurate meagurements of the relative volatilities were called for. 229 ORNL-DWG 66-3933 /‘f = ( ( NICKEL } TO VACUUM ) PIPE PUMP « D / q 5 3/8—1'n. 2-1n. (””’ NICKEL NICKEL PIPE | ) TUBING « « L BOILING LIQUID —. _ Hf//fC%PDENiFD T~ P AMPLE \\ THERMOWELL~_ DA AIR-WATER MIXTURE NICKEL TUBING Fig., 2.2. Diagram of Reciraulating Equilibrium S5till. A disgram of the recirculating equilibrium still used in recent work is shown in Fig. 9.2. The boiling section is a 12~in. length of 2~in.~diam nickel pipe. The condensing sectlon is made from l-in. nickel pipe wrapped with cooling coils of l/4-in. nickel tubing. In the bottom of the condenser is a condensatbte trap where liguid collects and overflows a welr to return to the still pot. The vacuum pump is connected near the bobttom of the condenser section. After charging salt of known composition, the still is welded shut and purged with argon. The desired pressure 1ls set, and the still is heated to the desired temperature while cocoling the condenser. After operating the stiil for a period of time at the selected conditions, the still is pressurized with argon, cooled to room temperature, and cub apart for examination and sampling. The concentrations of the rare-earth fluorides in the condensate trap and in the still pot are used to calcu- late relative volatilities according to Eg. (2). 230 In these experiments the determination of absolute values for rela- tive volatilities has been hampered because the vapor samples have rare- earth concentrations below the apalytical limit of detection. However, upper and lower limits of the relative volatilities for 0.01 to 0.02 mole fraction CeF3, NdF3, and LaFj in LiF al 1000°C and 0.5 mm Hg were deter- mined and are given below: < 0.0013 < 0y ;o < 0.0029 0.00055 <:OfidF3~LiF < 0.00089 = 0.00069, These values are substanhially lower than those previously reported. The samples have been submitted for analysis by a more sensitive analytical method (neutron activation). Higher concentrations of rare-earth fluo- ride will be used in the sbtill pot in future work in ovrder that rare- earth fluoride concentrations in the wvapor phase will be higher than the 1imit of detection. Thne importance of relative volatility in determining the operating characteristics of the distillation system is shown by the following cal- culation. Consilder the reboiler of a single-stage distillation system which contains V moles of LiF at any time and a gquantity of RBeFs such that vapor in equililibrium with the liquid has the composition of MSBR fuel salt. Assume that MSBR fuel salt containing Xg moles of rare~earth fluorides (REF) per mole of LiF is fed to the still pot at a rate of F moles of LiF per unit tilme where it mixes with the liquid in the system, Let the initial REF concentration in the liguid be Xg moles of REF per mole of LiF, and let the concentration at any time © be X moles of REF per mole of LiI'. From a material bhalance on RER, & (VX) = FXq ~ FO , (1) where V = still liquid holdup, moles of LiF, X = BEF concentration in still liquid, moles of REF per mole of Liy, I' = LiF feed rate to still, moles per unit time, Xo = REF concentration in feed, moles of REF per mole of LiF, X = relative volatility of REF referred to LiF. This equation nas the solution x =20 L - (1-a) A (2) The total quantity of REF fed to the system at time t is (Fb + V)X, and 231 the gquantity of REF remaining in the ligulid at that time is VX. Thus the fraction of the REF not vaporized at time t is e o vx___ {1/e) [L - o - /Y] () REF = (Ft + V)Xo — 1+ Ft/V) : Values for the fraction of REF retained in the still as a Tunction of dimensionless throughput (Ft/V) are given in Fig. 9.3 for various values of O, Approximately 91% of the REF will be retained in the still when 99.5% of the LiF has been recovered 1f the relative volatility of the REF is 0.00Ll; a retention of greater than 95% can be obtained for the same LiF recovery if & has a value of 0.0005. ORNL-DWG 66-11472 ] 2:0.0005 0001 FRACTION OF RARE-EARTH FLUORIDE RETAINED IN STILL 0 50 100 150 200 250 Ft/V, OIMENSIONLESS STiLL THROUGHPUT Pig. 9.3. Fraction of Rare-Earth Fluoride Retained in otil1. 232 9.2 Continuous Fluorination of a Molten Salt L. E. McNeege Uranium present in the fuel stream of an MSER must be removed prior to the distillation step since UF, present in the still would not ve completely volatilized and would in part be discharged to waste when the atill contents are dumped pericdically. Eduipment 1s being developed for the continuous removal of Ury from the fuel stream of an MSBR by contacting the salt with Fp in a salt-phase-continuocus system. This egquipment will be protected from corrosion by freezing a layer of salt on the vessel wall; the heat necessary for maintaining molten salt ad- Jacent to frozen salt will be provided by the decay of fission products in the fuel stream. Present development work consisgts of two parts: (1) studies in a continuous fluorinator not protected by a frozen wall, and (2) study of a frozen-wall system suitable for continuous fluori- nation but with which an inert gas is used. Experimental work on the nonprotected system is well under way; the protected system is being designed. The nonprotected system consists of a l-in.-diam nickel fluorinator 72 in. long and auxiliary eguipment (Fig. 9.4) which allows the counter- current contact of a molten salt with ¥Fy. Experiments can be carried out with molten-salt flow rates of 3 to 50 em /min with fluorinator salt depths of 12 to 54 in. The system is constructed of nickel with the ex- ception of molten-salt transfer lines, which are Hastelloy N. The flu- orinator off-gas passes through a 400°C NaF bed for removal of chromium ORMNL-DWG 65-93544 CHROMATOGRAPH FOR UF;, F, AND N, ANALYSES 1 METERED F, ~ | S0DA LIME TRAP 7 kg METERED N, FOR SALT DISPLACEMENT SALT SAMPLING ] A VESSEL gE - 1.5-in-DIAM OFF - GAS SALT RECEIVER SALT FEED TANK | - 14 liters 14 liters NICKEL FLUORINATOR {in. DIAM T2in. LONG Fig. 9.4, Equipment for Removal of Uranium from Molten Salt by Continuous Fluorination. 233 fluorides, a 100°C NaF trap for removal of UFg, and a soda~lime bed for Fo disposal. Analysis for Fp, Ufg, and Ny 1s made prior to the 100°C Nal' bed with a gas chromatograph. Iixperiments have been carried out at 600 to 650°C using an Nal'-LiF- 7Zr¥, mixture containing 0.2 to 0.5 wt % UF; and having a melting point of ~550°C. Salt feed rates of 5 to 21 cm?/min and Fu rates of 75 to 250 em®/min (STP) have been used with molten-salt depths of 40 to 52 in. Uranium removal during one pass through the fluorinator has varied from 95% to 99.6% as determined from salt samples. FEguipment operation has been smooth although several plugs have developed in salt transfer lines and in the fluorinator off-gas,. During the best run to date (CF—lO), a molten-salt feed rate of 20.7 cmfi/min and an ¥, feed rate of 250 em?® /min {STP) were maintained for a 2.5~hr period. The concentration of U in the feed salt was 0,32 wt %, and the fluorinator was operated at 650°C with a salt depth of 48 in. The U concentration in the salt discharged from the fluorinator during the last hour of operation was 0.0020 wt % zs determined from four salt samples taken at 15-min intervals. Based on inlet and exit U concentrations in the salt, 99,4% of the U was removed by the flu- orinator. The equipment operated smoothly during the run, salt and gas feed rates were constant, and the system appeared to have reached sieady state during the last hour of operation. Continuous fluorination of the fuel stream of an MSBR with equipment of the type studied is considered feasible. Study of continuous f{lucri- nation will be continued, and the use of a frozen wall for corrosion pro- tection will be demonstrated. 9.3 Alternative Chemical Procesesing Methods for an MSBR . I. Cathers Ce E. Bchilling Liguid-metal extraction was studied in a previous search for a salti reprocessing method to replace vacuum ddstillation. These tests with LisBel'y salt showed that the lanthanides were removed in most cases by a reductive coprecipitation with Be to yield refractory insoluble beryl- lides deposited at the salt-metal interface. The present problem was to develop a reductive precipitation using ninimum quantities of L1 or Be as reductant and involving simple physical separation of the precipitated metals by filtration or settling. Deconbtamination factors were determined wsing initial concentrations of 30 to 1000 ppm of Zr, Nd, Ia, Sm, Fu, Gd, and Sr either singly or in various combinations. Further studies on the liguid-metal extraction method were made using Li-Bi alloys to test removal of the same elements with the exceoption of Zr and N4, Table 2.1. Reductlve Precipitation of Neutron Poisons from LipBeF, Initial concentrations: Zr = 70 pom; Nd = 400 ppm; others £00-900 ppm _ ppm spike in original salt Factor of r T : - T s . ; o 2 - _T_ _ef. . ; = Ixperiment Reductant Fxcess Tempgrata*e ppm spike in treated salt Speg1§§ denti 1e9 in Met l. Nunber Reductant (°c) Precipitates by X-Ray Analysisg Zr La Nd Gd 1 Be 1 550-60C 2o 2.3 Be, 0-Zr (trzcel}, and TaBeys (trace) 2 Be 4.2 (NA) 550-60 55 2.0 Not examined 243 (Zr) 3 Be oty 550-95 2.0 8.0 TaBe1s sud Be (trace) 4 Be 103 550-80C 52 &-Zr, ZrBes, Be, and NdBejs N 5 Be 12 55060 395 Be > 6 Be 104 800 36 Cu (from filter) + unidentified Lines 7 Be 162 1000 1% Yot examined g Li 1.28 55075 234° 1.1 1.3 o-Zr; LiF + LiyBeF, in salt 9 Li 105 550—80 51 Cu (from filter) 10 Li 4 550 i3 6 Be, LaBe1s, FeBes (from crucible)” rocess required I 's 1.43 1.7 20 1.05 “Tnitial concentration Initial concentration separation by settliing. 400 ppm Zr, 11.7% Zr. Reduction Frecipitation The reduction-coprecipitation studies were made on 15-g (7.5-cc) samples of LipBeF, spiked with cold neutron poiscns as the fluorides; the samples were contained in mild steel under a protective abmosphere of argon., A typical test consisted of: preparation and sampling of the original salt while molten; reduction with between 1 and 243 times the theoretical amount of Li or Be ab temperatures in the range 500 to 1000°C for about 90 min with stirring by an argon sparge; filtration through a sintered Cu filter stick; recovery of frozen samples of treated salt from the filter and, when possible, precipitated metals from the bottom of the reactor. oalt samples were analyzed for lanthanides and zirconium by neu- tron actlvation analysis, spark source mass spectrometry, or emission spec- trometry. The metal precipitates were examined by x-~ray diffraction. The decontamination factors DF (La) and DF (Gd) shown in Table 9.1, when compared with the indicated process requirement at the bottom of each column, show that adequate removals could be achieved with both Id and Be using excess reductant in the range of 1 to 4 times the theoretical reguirement for LnBeis deposition. The Llow values obtained in experiment 8 reflect the incomplete removal of Zr, present in this one experiment as a major component (11‘7%). In this case the final concentration of Zr {500 ppm) was comparable to the initial concentrations of Is and Gd, and therefore sufficient to inhibit their reduction. The species iden- tified in the metal precipitates recovered {see Table 9.1, experiments 1, 3, and 10) indicate that reduction of lanthanides in LisBel, with elther L1 or Be leads exclusively to berylilides of the type ImBeis. The x-ray identifications listed in Table 9.) represent only the most likely specific beryllide since all the InBejs compounds of the lanthanide series are isostructural, with very nearly identical lattice parameters, and therefore are indistinguishable from each other when codeposited from = mixture. Fxamination of zirconium DF's shown in Table 2.1 show it to be re- moved easlly from solutions in LizBelF, over a broad range of concentra- tion by treatment with elther Ii or Be. It ig deposited elther as free metal or from very dilute solutions (70 ppm) as ZrBe, (Table 9.1, experi- ment 4). The value of DF (Zr) = 11 obtained at 1000°C in experiment 7 is of special interest since it suggests a solution to the high volatility of Zr¥F,, a major problem encountered at this temperature in the vacuum- distillation salt recovery method. Lithium or beryllium could be used to retain Zr in the still pot as nonvolatile free metal. In experiment 4, due to the low zirconium concentration (70 ppm), ZrBe, was ldentified along with G-Zr., In addition to the elements listed in Table 9.1, Sm, Bu, and Sr were studied as parts of the mixtures used in experiments 1, 3, and 8§ plus other tests not listed. Aside from a IF (Sm) = 2 obzerved in ex- periment 3 and a DF (Sr) = 1.14 in an unlisted test uging Idi at 550°, no encouraging resulbs were obtained by this method for these elements. Tabie 9.2. Li-Bi Alloy Extraction of Lanthanides from Molten LipBelly Tritial concentration: lsnthanides = 800-L000 ppm each; Sr ~200 ppm Contacting Conditions C . - _ original sall mole fraction in metal Ailoy Composition = Ctre"ted et KD T mole fraction in salb (at. %) Temperature Tlme Volume Ratio = mEs °n ) e ) - 3 (“c. (win) (alloy/sale) .= g A Ge St e Sm s ad Sr 5,0 Li—94.1 Bi 55060 160 0.71 3.0 3.5 1.0 2.0 1.5 0.85 2.2 0.5 0.32 0.18 (C.. . =2.56 at. % Li) final 30 Li-=70 Bi 55060 120 .l 6.0 2.0 8.0 5.0 2 1.1 3.1 17.1 2.00 Q.97 (C.. = 10.05 at. % Li) final DE reguired 1.7 5.0 1.43 1.05 1.04 9¢c 237 Li-Bl Alloy Extraciion Liguid-metal extraction tests with sclutions of ILi in Bi were made in the same equipment but without filtration of either phase. Samples were recovered after settling, quenching to room temperature, and sac- rificing the crucible. Both salt and metal slugs were cleaned of sur- face deposits before sampling for analysis. The decontamination factors (DF's) and distribution coefficients (KD) obtained in liquid-metal extractions of spiked carrier salt using Ii-Bi alloys are summarized in Table 2.2. TFor the spectrum of spikes present (La, Sm, Bu, Gd, and Sr) adequate decontamination was achieved for all but Sm. The generally high DF's cannot be explained in most cases by a true extraction mechanism, a fact indicated by the low dis- tribution coefficlents shown in Table 9.2. Also, poor material balances were obtained through use of analytical results from samples of the salt and metal phases. The explanation for removal of the elements in question ig precipitation as interfacial solids which were later identified by x-ray analysis to be beryllides of the InBeps type. The high Kj observed for Fu indicates that this element was removed almost exclusively by ex- traction when using 30 at. $ Li-Bi alloy. References 1. C. D. Scott and W, L. Carter, Preliminary Design Study of a Continuocus Mluerination ~ Vacuum-Distillation System for Regenerating Fuel and Fertile Streams in a Molten Salt Breeder Reactor, ORNL-3791 (January 1966). 2. MSR Program Semiann., Progr. Rept. Feb. 28, 1966, ORNL~3936, p. 199. 239 OAK RIDGE NATIONAL LABORATORY MOLTEN-SALT REAGCTOR PROGRAM JUNE 15, 1965 MRC MEC MEC M&C M&C MBC M3.C Ma.C M&C M&C M&C M&C M&C M&C R. 8. BRIGGS, DIRECTOR D P. R. KASTEN, DEPUTY DIRECTOR R : METALLURGY MSBR DESIGN STUDIES MSRE OPERATIONS COMPONENT AND SYSTEMS DEVELOPMENT JAC DESIGN AND DEVELOPMENT REACTOR CHEMISTRY E. S BETTIS R W. R. GRIMES* RC G. M. ADAMSON* o ‘ = P. N. HAUSENREICA R DUNLAP 5COTT R R. L. MCGRE 1&C F. F. BLANKENSHIP RC W. H, COOK** W. L. CARTER* cT £ G. B0HLMANN* RC B A CANONICO® A. G. GRINDELL* R H. F. McDUFFIE* RC o -G - E. A. FRANCO-FERREIRA H. T. KERR R : 2 C ROBERTSON R MSRE ON-SITE CHEMISTRY C. R. KENNEDY L [ = H. E. McCOY* W. TERRY S. 5 XIRSLIS RC 1. L. GRIFFiTH* ). K. JONES R . P. D. NEUMANN® RC V. G. LANE" COORDINATION NUCL EAR AND MECHANICAL ANALYSIS ENGINEERING ANALY SIS A.H. ANDERSON I&C E J LAWRENCE® D. G. DAVIS 1&c IRRADIATION PROGRAM R G SHOOSTER® C.H. GABBARD"* R J. R. ENGEL R 2 1 KEDL a P. G, HERNDON 18C E M. 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