The importance of effective waste management in the nuclear fuel cycle cannot be overestimated. At Oak Ridge National Laboratory (ORNL), development work in waste reprocessing and treatment includes the testing and use of various additives for the purpose of facilitating adherence to both process and regulatory performance criteria. Three waste reprocessing/treatment technologies and the associated materials are discussed in this paper: (1) suspension and transfer of sludge from waste storage tanks; (2) treatment to render a waste in compliance with regulatory requirements; and (3) fluoride-rich waste reprocessing. 7 refs., 3 figs

Delzer, D. B.

McDaniel, E. W.

Sams, T. L.

Tallent, O. K.

Weeren, H. O.

UNT Digital Library

emf'&wscf—lMATERIALS USED IN LOW-LEVEL LIQUID WASTE REPROCESSING/TREATMENTSTUDIES AT OAK RIDGE NATIONAL LABORATORY*c » g 3 g | gg g)H CONF-870801—1I^llfi-iff la E- W- McDaniel+S?l: | 8 I. Ss s | 1 1 H. 0. Weeren DE87 010601| I o 3 ~ I ? | r | D. B. Delzerg>2>s>g g i § | z | T. L. Sams3 S I S » o 3 » 53 0. K. TallentJ | £ §• | Sf-1 § g Chemical Technology Division> < | c ^ 3 5 . S ^ i . g Oak Ridge National LaboratoryI H|B .Nsi « Oak Ridge, Tennessee 37831| 8 - * i ss | l ' l l rI S o 8 1' 3i 1 1 ̂  To be Presented at| « 3- J | | 1 1 | The Conference ono 1 »• ~ Jl-a iS" g. | Nuclear Fuel Reprocessing and Waste Management (RECOD 87)!^sh?-'° Paris, Frances t l l l l l S i A u g« s t 2*"28, 1987» 2 , a o - S ^ S c and Published in the ProceedingstiHr*ln"Ttx vjbminad manuaciipt h u M*nauttior«d by a contraclor of I t * U.S.Gov*mm«nt unctor contract No. DE-AtX«-840»«21400. Accordngly. I t * U.S.Gov*rnm«o! r«t«ins • nonanduMV*.roy«)tY-1r«« bc«OM to publiah or raproduoarho puokshtd lixni o< ttiiv contribution, ordow othart to do to. lor U.S. GovtrramntpurpOMS.'•Research sponsored by the Office of Defense Waste and ByproductsManagement, U.S. Department of Energy, under contract DE-AC05-840R2L400with Martin Marietta Energy Systems, Inc., Oak Ridge, Tennessee, U.S.A.+Denotes speaker.MASTER ^DISTRIBUTION Of THIS 08CWTCN1 * W t l M i mMATERIALS USED IN LOW-LEVEL LIQUID WASTE REPROCESSING/TREATMENTSTUDIES AT OAK RIDGE NATIONAL LABORATORYSFEN-RECOD 87/0045/JG/daE. W. McDaniel, H. 0. Weeren, D. B. Delzer, T. L. Sams, 0. K. Tallent[Chemical Technology Division, Oak Ridge National Laboratory,Oak Ridge,--Tennessee U.S.A. ]ABSTRACTThe importance of effective waste management in the nuclear fuel cyclecannot be overestimated. At Oak Ridge National Laboratory (ORNL),development work in waste reprocessing and treatment includes, the testingand use of various additives for the purpose of facilitating adherence toboth process and regulatory performance criteria.Three waste reprocessing/treatment technologies and the associatedmaterials will be discussed in this paper: (1) suspension and transferof sludge from waste storage tanks; (2) treatment to render a waste incompliance with regulatory requirements; and (3) fluoride-rich wastereprocessing.INTRODUCTIONDuring the past 44 years, the Chemical Technology Division (CTD) of ORNLhas been a major contributor to all aspects of the nuclear fuel cyclein the defense and commercial sectors. Early development emphasized fuelprocessing and reprocessing technology. In recent years, the emphasis hasshifted to the "back end of the fuel cycle" or waste management.This paper discusses three waste reprocessing/treataent technologiesto which CTD has contributed in recent years and the materials associatedwith these technologies. The technologies discussed are (1) suspensionand transfer of sludge from waste storage tanks; (2) treatment to render amixed waste in compliance with regulatory requirements; and (3) fluoride-rich waste reprocessing. In each case, the addition of a specificmaterial will result in the desired properties.SLUDGE REPROCESSING :In 1943, the Clinton Laboratories were built at Oak Ridge, Tennessee, toserve as a pilot plant for production operations at Hanford, Washington.The work Included the construction of six underground tanks for storage ofthe wastes generated at the laboratories. These tanks were in operationaluse by November 1943. Subsequently, Clinton Laboratories became Oak RidgeNational Laboratory, and the waste storage tanks became an integral partof the laboratory's waste system. The waste handled by this system isroutinely treated with caustic to a pH of 10 or greater; substances thatare insoluble in alkaline solutions precipitate and settle, principally inthe waste storage tanks. By 1980, when the waste system was revised and(1)these tanks were no longer being used, they had accumulated-1.5 x 106 L (400,000 gal) of sludge containing the insolublefraction of the radionuclides that had been generated at the laboratory.A program to empty the waste tanks and dispose of the accumulatedradioactive sludge had already begun (in 1977). Process development,system design, and facility construction were completed by June 1982.In ~18 months of operations, all six tanks were sluiced and ~90Z of thesludge was resuspended1 and was pumped to the 0RNL Hydrofracture Facilityfor temporary storage and, subsequently, permanent disposal.2The diameter of the waste storage tanks is 15 m (50 ft) with a centerdepth of 5.5 m (18.25 ft) and an individual tank capacity of ~640,000 L(170,000 gal). The tanks were built by spraying a cement slurry againsta lattice of reinforcing bars (the Gunite process) for a final wallthickness of 19 cm (7.5 in). A French drain at the base of the outsidewall was provided with a leak detection system, but there was noprovision for double containment. The interiors of the tanks arerelatively uncluttered; there is a pump suction leg and ?. level probe,but no cooling coils.The depth of sludge varied from ~30 cm (1 ft) found in three tanks to3 m (10 ft) in one of the tanks. Chemical and radiochemical analyses ofsludge samples showed great variability among the tanks, between differentlevels in the same tank, and even between supposedly duplicate samplestaken from the same level in the tank. Data collecting through inventorysampling showed the major elements in the tanks to be uranium, iron,thorium, calcium, and aluminum. The primary radionuclide (representing95% of the total activity) was 90Sr. Analyses indicated ~2 x 106 Ct of90Sr (which was later shown to be an overestimate) and 4,500 Ci of TRUisotopes (about 10 times higher than estimates based on accountabilityrecords). Radionuclide assays subsequently done in support of disposaloperations indicated ~73O,OOO Ci of 90Sr and 2,500 Ci of TRU isotopes.These figures are about half the estimates generated by grab samplingprior to sluicing. The lower figures are considered to be more accuratebecause they represent measurements made on samples from the "homogenized"sludge.Laboratory tests indicated that ~50 wt X of the sludge consisted ofsmall particles (< 10 Mm diam) which settled slowly in water. Tests ona separated fraction of the particles > 10 um diam showed that they werequite friable. A series of laboratory and field tests with simulatedsludges and hot-cell tests with actual sludges demonstrated3 that:1. the sludge agglomerates could be reduced to < 20-Mm size with acommercially available grinder;2. sludge particles < 20 Mm diam would not settle for severaldays if dispersed in 2.5 wt % bentonite suspension;3. slurries containing 2.5 wt % bentonite could be pumped atconcentrations up to 25Z wt Z;4. a single sluicer mounted high in each tank could be used toresuspend the sludge;5. instrumentation was a problem because most available instrumentswould not work on slurries under field conditions; and(2)6. vision into the tank during sluicing would be essential tomonitor process operations.A diagram of the batch process used for emptying the waste tanks isshown in Fig. I. In this process, a bentonite suspension is continuouslypumped between the tank being sluiced and a near-empty waste tank locatednearby. Initially, a 150,000-L (40,000-gal) batch of 2.5Z bentonitesuspension in water is mixed and collected in the near-empty tank thatserves both as a feed tank and as the container for the resuspended wastesludge. This suspension is then pumped through a remotely controlledsluicer nozzle to impinge on the sludge in the waste tank being emptied.The Impact of the jet stream breaks up the settled sludge and resuspendsthe individual particles while the bentonite suspension hinders settling.The resuspended sludge is pumped from the waste tank through a grinderthat breaks up oversized particles and then returned to the feed tank.This operation continues until the sludge concentration in the feed tankapproaches 15 wt Z. At this point, the sluicing is stopped, and theslurry in the feed tank is pumped to storage. A new batch of bentonitefeed suspension is then prepared, and the cycle is repeated until thewaste tank is judged empty. Fig. 2 shows the arrangement of tank sluicingequipment.Calculations indicated that the weight of the sludge solidstransferred from the six waste storage tanks was ~40% more thanwas originally thought to be there. Radionuclide assays were aspreviously described. Each of these values Is approximately half thevalue estimated from analyses of grab samples prior to the sluicingoperations.In evaluating possible suspending materials, a delicate balance between(1) having enough material to suspend the sluiced sludge and (2) creatinga gelatinous mass that could not be pumped, had to be maintained. Manyorganic viscosity "enhancers" were evaluated that fell into the lattercategory. Others were rejected because of the uncertainty In thelong-term effect of the waste stream on the suspender.Laboratory experiments determined that both bentonite and attapulglteclays were effective suspending agents. Bentonite was chosen because ofmore reliable performance in laboratory settling tests and viscositydeterminations.^REPROCESSING/TREATMENT IN ORDER TO CONFORM TO REGULATORY REQUIREMENTSIn the United States three categories of radioactive waste exist:(1) high-level or that which has had direct contact with the fuel element;(2) Transuranic (TRU), if TRU isotopes equal or exceed 100 nCj/gram; and(3) low-level. These wastes are regulated by either U.S. Department ofEnergy (USDOE) or Nuclear Regulatory Commission (NRC) directives.The disposal of low-level wastes in grout (a cement mixture) requires, ofcourse, that the leaching of radionuclides be kept witnin acceptablelimits, as well as that the waste form meet standards of compressive(3)ORNL DWG 1 2 - 1 0 5 3WATERWATERPREPARATIONTRANSFER ;Fig. 1. Schematic of batch process used for emptying the waste tanks.0*1*1. owe n ICOUIMICNT ENCLOSURES <TVflC*Cv MIXING TANK IM 10 \-TYMCAl TANKS « 5. W«. •» ». «*. HI. ANO W HFig. 2. Arrangement of tank sluicing equipment.(4)strength and phase separation. In other words, the radionuclldes shouldbe retained in the set grout and not lost either in the liquid phase, ifphase separation occurs, or by leaching from the solid after the grout hasset.Non-gelling clays were evaluated for their ion-exchange capabilitiesin studies at ORNL. These illitic clays serve as a natural ion exchangerfor cesium; their inclusion in a grout mix greatly improves ^Csretention. Retention of radioactive strontium was shown in tests atORNL to be relatively unaffected by the mix ratio (proportion of waterto cement).The amount of calcium in the mix did have an effect, and Portland cementis composed mainly of calcium silicates, which react with water to formhydrated calcium silicates. Since strontiuc and calcium have similarchemical behavior, 90Sr rapidly enters into these reactions in common withcalcium. Reducing the amount of calcium in the waste immobilizationmatrix reduces the strontium leachability by increasing the availabilityof reaction sites for strontium.Suspension of the solids, and thereby reduction of phase separation, isan important objective of grout formulation. If an appropriateminimization of phase separation can be achieved, the waste loading canbe increased (within limits) which makes the process more economical. Itwas found that a gel clay additive can be used in this way to optimizewaste loading.1*'5If the materials discussed above are blended with cements in the properamounts and mixed with the waste, the result is a grout that can beprocessed in a variety of equipment and, if permitted to solidifybefore disposal, gives a product that will meet or exceed EP-TOX, USDOE,and NRC regulations.6TREATMENT OF FLUORIDE-RICH WASTEIn fabricating nuclear fuel elements, metallurgists have found thatzirconium alloys make excellent fuel claddings for selected fuel mixtures.The chop-leach method is used to reprocess the irradiated fuelassemblies whereby the fuel is removed by treating the hull with anammonium fluoride solution. The hulls and other fuel hardware arecategorized as solid waste and are disposed of in an appropriate mannerwhich is outside the scope of discussion in this paper. The leached fuelis separated from the leachate which contains little of the fuel butsignificant amounts of dissolved zirconium. This fluoride-rich liquid,which by USDOE definition is high-level waste, must be further processedto render it suitable for storage and perhaps disposal. The slightlyacidic liquid's pH is increased to 12 or 13 by the addition of sodiumhydroxide to render it suitable for storage in carbon steel tanks. Anydissolved zirconium precipitates, along with sodium flouride, at high pH.(5)The resulting filtrate is little more than water and is separated andtreated as waste water. The precipitate, which may be either TRU orlow-level waste, is ready for disposal after a suitable cooling period. Aflow diagram is shown in Fig. 3. However, the cladding removal wasteprecipitate, presents several problems, with the major one being fluoride,which is difficult and expensive to vitrify in glass. If a cement-basedhost is chosen, fluoride prevents cement from setting. Furtherreprocessing, such as washing, serves only to dilute the fluoride. Anobvious solution is a combination treatment-disposal step that renders thefluoride, chemically benign, thereby making it possible to immobilize theslurry with only sluicing and transfer to an immobilization system.Studies at Oak Ridge National Laboratory have developed a process thatovercomes the retardation problem when a cement-based matrix is usedby adding Ca(0H)2 to the matrix-forming materials which reacts with thefluoride to produce insoluble CaF2. The reaction proceeds according tothe following equation:Ca(OH)2 + 2NaF •*• CaF2r + 2NaOH.Thus, the addition of concentrated NaOH and Ca(0H)2 serves to immobilizethe zirconium and fluoride in the waste. Other waste constituents areimmobilized by the physical characteristics of the grout.Preliminary experimentation confirmed that the fluoride was indeedbeing inactivated by the calcium hydroxide. The results of these testsled to the development of an empirical equation used to optimize theamount of calcium hydroxide added to the dry—solids blend. The followingequation defines the fluoride/calcium equivalence ratio in a neutralizedcladding removal waste grout:0.2125 DsEq = _ _ _ _ _ = 1(Ca2+) PRwhereEq « fluoride/calcium equivalence ratio;Dg •* sludge dilution factor;P = Ca(0H)2 in dry blend, wt Z;R = mix ratio, kg/L.Based on this equation, grouts with preferred properties should beobtained when the equivalence ratio is ~1.0 which was found to be true,7and the requirement of Eq < 1 was subsequently used to determine thepreliminary reference grout formula shown below:(6)ORNL DWG 87-530RFUEL HULLSAMMONIUM FLUORIDE SOLUTIONFLUORIDESOLUTIONpH 6.0HULLSSOLID WASTEDISPOSALLIQUID TOEVAPORATORSLURRYOFFLUORIDESALTSSEPARATIONANDREPROCESSINGCa(OH)2TREATMENTFig. 3. Flow diagram of the chop-leach method for reprocessing fuel assemblies.Type III Portland cement 45 wt %Centralla, Washington, Class F fly ash 35 wt ZHydrated lime [Ca(0H)2J 12 wt %Indian Red.Pottery clay 8 wt %Mix Ratio ' . ; . 0.84 kg/LThis grout formula, which is the result of many experiments, became thebasis for the final formulation experimental design.CONCLUSIONSA variety of process technologies are available to treat and processdifferent waste types to render them suitable for immobilization anddisposal.Oak Ridge National Laboratory has developed technology and equipment forsludge removal and transfer, treatment of mixed waste, and technology forimmobilizing fluoride-rich waste.REFERENCES1. H. 0. Weeren, L. C. Lasher, E. W. McDaniel, "Cleanout of Waste StorageTanks at Oak Ridge National Laboratory," Materials Research SocietySymposium Proceedings, Boston, Massachusetts, November 26-29, 1984,Vol. 44, pp. 785-792.2. Disposal of Radioactive Grouts into Bydraullcally fractured Shale,Technical Reports Series No. 232 IAEA, Vienna, 1983.3. H. 0. Weeren and T. S. Mackey, Waste Sludge Resuspension and Transfer- Development Program, Oak Ridge National Laboratory, ORNL/TM-7125,1979.4. W. de Laguna, T. Tamura, H. 0. Weeren, E. G. Struxness, W. C. McClain,and R. C. Sexton, Engineering Development of Hydraulic Fracturing asa Method for Permanent Disposal of Radioactive Wastes, 0RNL-4259,August 1968.5. J. C. Moore, H. W. Godbee, A. H. Klbbey, and D. S. Joy, Developmentof Cementitious Grouts for the Incorporation of Radioactive Wastes.Part 1: Leach Studies, ORNL-4962, April 1975.6. T. M. Gilliam, J. A. Loflin, Leachability Studies of HydrofractureGrouts, ORNL/TM-9879, November, 1986.7. T. L. Sams, 0. K. Tallent, T. M. Gilllam, and E. W. McDaniel,Formulation Studies and Grout Development for Fixation of CladdingRemoval Waste, Milestone 181, 0RNL/TM-9810, Oak Ridge NationalLaboratory, Oak Ridge, Tennessee, in preparation.(8)

Materials used in low-level liquid waste reprocessing/treatment studies at Oak Ridge National Laboratory

https://digital.library.unt.edu/ark:/67531/metadc1211260/m2/1/high_res_d/6431292.pdf

Materials used in low-level liquid waste reprocessing/treatment studies at Oak Ridge National Laboratory

Abstract

Similar works

Full text

Available Versions

UNT Digital Library