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A. U.S. Department of Energy LaboratoryLALP-93-92

POLLUTIONPREVENTIONTECHNOLOGIESFORPLUTONIUMPROCESSING

NUCLEARMATERIALSTECHNOLOGY1994

Published byLos Alamos National LaboratoryNuclear Materials Technology DivisionMS E500Los Alamos, New Mexico 87545

CoordinatorK.K.S. Pillay

EditorsCarolyn K. RobinsonJoan Farnum

Graphic DesignerSusan L. Carlson

Printing CoordinationGuadalupe D. ArchuletaBrenda R. Valdez

For More InformationK.K.S. Pillay(505) 667-5428

Page 65 Photos:figure 3—B 1420 01figure 4—B 1425 15

Los Alamos National Laboratory is operated by the University of California forthe U.S. Department of Energy under contract W-7405-ENG-36.

An Affirmative Action/Equal Opportunity Employer.

This report was prepared as an account of work sponsored by the United StatesGovernment. Neither the Regents of the University of California, the UnitedStates Government, nor any of their employees makes any warranty, expressedor implied, or assumes legal liability or responsibility for the accuracy, com-pleteness, or usefulness of any information, apparatus, product, or processdisclosed, or represents that its use would not infringe privately owned rights.Reference herein to any specific commercial product, process, or service bytrade name, mark, manufacturer, or otherwise does not necessarily constituteor imply its endorsement, recommendation, or favoring by the Regents of theUniversity of California, the United States Government or any agency thereof.The views and opinions expressed herein do not necessarily state or reflect thoseof the Regents of the University of California, the United States Government, orany agency thereof.

LALP-93-92April 1994

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Contents

C o n t e n t s

PrefacePreventing Pollution From Plutonium Processingby K.K.S. Pillay 3

Overviews - Waste Minimization TechnologiesOff-Gas Scrubbingby W. Brad Smith 7

Acid Recovery and Recycleby Thomas R. Mills 12

Hydrochloric Acid Waste Stream Polishingby Louis D. Schulte, Richard R. Salazar, Benjie T. Martinez,John R. FitzPatrick, and Bradley S. Schake 16

A Novel Liquid-Liquid Extraction Techniqueby Daniel J.Kathios, Gordon D. Jarvinen, Stephen L. Yarbro,and Barbara F. Smith 20

Electrochemical Treatment of Mixed Hazardous Wasteby Wayne H. Smith, Christine Zawodzinski,Kenneth R. Martinez, and Aaron L. Swanson 37

Application of Freeze-Drying Technology toDecontamination of Radioactive Liquidsby Nicholas V. Coppa 44

Nuclear Applications for Magnetic Separationsby Larry R. Avens, et al. 52

Distillation Separation of Actinides from Waste Saltsby Eduardo Garcia, Vonda R. Dole, Walter J. Griego,John J. Lovato, James A. McNeese, and Keith Axler 61

Evaluating Existing Partitioning Agents for ProcessingHanford High-Level Waste Solutionsby S. Fredric Marsh, Zita V. Svitra, and Scott M. Bowen 66

Gas Pycnometry for Density Determination ofPlutonium Partsby J. David Olivas, S. Dale Soderquist, Henry W. Randolph,Susan L. Collins, Darren Verebelyi, William J. Randall,and Gregory D. Teese 71

Contents continued on next page

allan.m

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Nuclear Materials Technology 1994 Report2

Contents

C O N T E N T S Continued

Plutonium Dry Machiningby Mark R. Miller 75

Technical Issues in Plutonium Storage

by John M. Haschke and Joseph C. Martz 80

Overviews - Facility OperationsWaste Management at Technical Area 55: A HistoricalPerspective and Status Reportby Bill J. McKerley, James J. Balkey,and Ronald E.Wieneke 88

Future of Waste Management at the Los AlamosPlutonium Facilityby Ronald E. Wieneke 94

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Preface

K. K. S. PillayNuclear Materials Technology

P r e f a c e

Fig.1. Accumulation of environmental statutes and their amendments since the passage of the Atomic Energy Act in 1954.

IntroductionPollution is a dirty word. No one has to make a case anymore forthe need for pollution prevention or the related concept of environ-mental management. Although the history of pollution control leg-islation in the United States dates back to the River and Harbor Actof 1899, the past four decades have been most active in developingand administering numerous environmental regulations. This pro-cess of administering environmental regulations has accelerated in

recent years and has become acritical factor in the quality oflife in these United States.

Since the passage of the AtomicEnergy Act in 1954, the Congressof the United States has promul-gated numerous environmentalstatutes. Figure 1 shows the ac-cumulation of environmentalstatues and their amendmentsthat are presently applicable tonuclear defense facilities. Thesestatutes have been codified intoover 9000 regulations by federalagencies.1 The EnvironmentalProtection Agency (EPA) alonehas over 125,000 employeesadministering these statutesand regulations. There is anequal number of environmentalregulators at the state and localgovernment levels.

Although radioactive wasteshave been accumulating in theUnited States for the past halfcentury, only during the lasttwo decades has there beena concerted effort to addressthe problems of long-termisolation of these wastes fromthe human environment.After considerable study anddebate, the U.S. program forradioactive waste managementwas codified in the NuclearWaste Policy Act of 1982.2

Preventing Pollution From Plutonium Processing



Preface

According to this legislation,the U.S. Department of Energy(DOE) has the responsibilityto develop strategies, systems,and technologies for the long-term isolation of all spentnuclear fuels, high-level wastes,and transuranic wastes in theUnited States. Because of theunique socio-political atmo-sphere surrounding radioactivewaste management, there areconsiderable challenges to thescientific community for devel-oping technological solutionsthat are acceptable to aconcerned free society.

For nearly four decades, the U.S.nuclear defense facilities soughtrefuge under the Atomic EnergyAct and claimed sovereign im-munity from compliance withmost of the environmental regu-lations. In response to increasingpublic pressure during the pastfew years, the defense sector hasundergone a cultural shift thathas changed past attitudes to-ward complying with environ-mental regulations. This shifthas culminated in the passageof the 1992 Federal FacilityCompliance Act (FFCA) andother complementary regula-tions and executive orders.3,4

The FFCA gives federal facilitiesa three-year period to developdetailed plans for complyingwith all applicable environmen-tal regulations. This particularlegislation has been a greatmotivator to all federal facilities,including the Los Alamos Pluto-nium Processing Facility, tomodify past practices and tobecome conscious of the envi-ronmental impacts of all itsactivities.

The following paragraphsexamine the legacy of nuclearwaste, the new priority of wasteminimization, responsible wastemanagement, and technologicalchallenges for the future.

The Legacy of Nuclear WasteThe term “nuclear waste” isgenerally used to represent allradioactive waste materials re-gardless of their origin, materialform, or radioactivity levels.Nuclear wastes originate from amultitude of activities involvingnuclear technologies in boththe civilian and defense sectors.The steady accumulation ofnuclear wastes over the past halfcentury has attracted consider-able public attention during thepast two decades. Following thereactor accidents at Three MileIsland and Chernobyl, therehave been overwhelming ex-pressions of public concernsover safety of all nuclear facili-ties and the safe managementof accumulated nuclear wastesas well as those that are beinggenerated.

Studies have shown that thepublic’s perception of risk fromnuclear waste is similar to thoseof highly energetic, catastrophicevents. This perception isqualitatively incorrect becauseradioactive wastes have farlower energy levels in compari-son with reactor accidents,earthquakes, volcanic eruptions,etc. An educational process thatconveys realistic risks of nuclearwastes has to be a necessarypart of an overall program ofnuclear waste management.

The Nuclear Materials Technol-ogy (NMT) Division, in this newcultural climate, adopted a stra-tegic goal to systematicallyeliminate the adverse environ-mental impacts associated withoperating the Los Alamos Pluto-nium Facility. This ambitiousgoal of a committed group ofpeople in the Los Alamosnuclear technology communityis the theme of this compilationof technology initiatives. Ourstrategic goal is not only part ofa survival strategy but also rep-resents our readiness to solvenational environmental prob-lems associated with nuclearprocessing facilities and,thereby, help alleviate publicand governmental concerns.

Since September 1992, severalactivities within NMT Divisionhave been postponed becauseof the FFCA and the long andlaborious negotiations that arerequired to reach a facility-specific agreement with theEPA. On July 19, 1993, DOESecretary Hazel O’Leary ap-proved the Los Alamos agree-ment with Region 6 of the EPA.According to this agreement,full compliance with the FFCAis expected by the year 2000.During these negotiations,NMT Division began to imple-ment its strategic plan to meetall regulatory requirements ofwaste minimization and pollu-tion prevention and to transformthe Los Alamos plutonium pro-cessing operations into an envi-ronmentally benign activity.In developing those technolo-gies required to meet our goals,we will also serve the needs ofthe weapons complex of the nextcentury.

5

Preface


environment are completelyeliminated. The DOE has recog-nized that if it is to successfullyachieve its long-term goals, thisagency must first find a way tochange public perceptions aboutnuclear waste. Focusing onwaste minimization andresponsible waste managementare steps in the right direction.

Waste MinimizationWaste minimization is any ac-tion that avoids or reduces thegeneration of wastes. For nearlya decade, defense nuclear facili-ties have been voluntarily mak-ing efforts to reduce nuclearwaste generation. In recentyears, DOE identified environ-mental restoration and wastemanagement as a major missionarea.6 The primary objective ofthe fourth annual Environmen-tal Restoration and Waste Man-agement Five-Year Plan for1994-98 is the reduction ofwastes from DOE facilities.7

Waste reduction can be accom-plished through source reduc-tion, recycling potential wastematerials that cannot be elimi-nated or minimized, and treat-ing all wastes to reduce theirvolume, toxicity, or mobilityprior to storage or disposal.The significant reduction innuclear weapons requirementsrepresents a major source reduc-tion mechanism. Recycling ofa number of waste streamcomponents is a technical

Arguments about the costs andbenefits of nuclear technologiescontinue to dominate publicdiscussions. Attitudes of bothgovernmental agencies andnuclear industries based onartificial cost-benefit analyseshave contributed enormouslyto the negative attitudes sur-rounding nuclear wastes in theUnited States. Unfortunately,these analyses do not factor inthe significant costs arising fromenvironmental impacts consid-ered unacceptable by the generalpublic.

The unmaking of nuclear wastesand the abrogation of all uses ofnuclear technologies are not inthe realm of reality. As weapproach the 21st Century, wecannot afford to forgo the manypeaceful uses of nuclear tech-nologies. Considerable resourceshave been devoted during thepast two decades to the designof technical solutions for long-term management of nuclearwastes and isolation of thiswaste from the biosphere. Thereis a world-wide scientific con-sensus that deep geologic dis-posal is the best option fordisposing of existing inventoriesof high-level and transuranic ra-dioactive wastes.5 However, itis now abundantly clear that thepublic and its representativeswill not be satisfied until re-leases of radioactive nuclidesfrom man-made sources to the

possibility now and can accom-plish additional waste reduction.Waste treatment to reduce vol-ume, toxicity, and mobility are allin various stages of developmentand implementation.

DOE’s ambitious plan for wasteminimization can be taken a stepfurther using new technologiesessential to new nuclear facilitiesthat will have little or no impacton the environment. Liquidwaste-stream polishingto eliminate radioactive andhazardous components of wastesfrom nuclear materials process-ing facilities for the weaponsfuel cycle is feasible and can beimplemented in large-scalenuclear materials processingfacilities. Approaches to mini-mizing radioactive and hazard-ous components in solid wastestreams are being developed,and the costs for implementingsuch technologies are not ex-pected to be prohibitive. Becausea modern nuclear material pro-cessing facility is necessary tomaintain a nuclear deterrent ca-pability, a vigorous technologydevelopment program for wasteminimization is essential tomaintain public confidence innuclear programs and to changepresent attitudes toward nuclearwaste.



Preface

materials and its impact on ourfuture is an international politi-cal problem with no obvious so-lutions. Achievements in thearea of nuclear arms reductionare eclipsed by the increasinglikelihood of nuclear prolifera-tion. Although we can proposeto control proliferation throughsystematic reduction in fissilematerial inventories worldwide,the need for energy productionin many parts of the world usingfissile and fertile materials mustbe met.

These challenges are opportuni-ties for NMT Division to de-velop and apply technologiesthat can prevent environmentalpollution. Technology transferinitiatives now being activelypromoted by DOE offer oppor-tunities for NMT Division todevelop university and industrypartnerships that can be mutu-ally beneficial. Technology ini-tiatives detailed in the followingpages of this report are stepstoward achieving our overridinggoal of an environmentally be-nign plutonium processing facil-ity at Los Alamos.✦

Responsible Waste Manage-ment and Future ChallengesFuture nuclear processing facili-ties have a major responsibilityto properly treat and dispose oftheir wastes. These facilitiesmust be able to function withoutthe release of radioactive andhazardous materials to the envi-ronment. Dilution and disper-sion of wastes practiced in thepast are not acceptable solutionsany longer. Immobilization andisolation of wastes from the bio-sphere will have only a limitedrole in the long-term manage-ment of nuclear wastes. Reposi-tory construction and geologicdisposal as a solution to nuclearwaste problems cannot be con-tinued forever; however, stabili-zation and storage have to beseriously considered as an op-tion so that valuable resourcesare not discarded for politicalexpediency. Managed storageof concentrated nuclear materi-als may become the policy forthe national defense sector.

The scientific community facesnumerous other challenges re-sulting in part from present pub-lic perceptions of nucleartechnologies. These challengesinclude meeting the conflictingneeds of nuclear proliferationcontrol and nuclear energy pro-duction. Increasing concernsover the proliferation of nuclear

Cited References1. Philip H. Abelson, “PathologicalGrowth of Regulations,” editorial,Science 260, 1859 (June 25, 1993).

2. U.S. Congress, “Nuclear WastePolicy Act of 1982, ” Public Law 97-425, 97th Congress (January 1983).

3. U.S. Congress, “Federal FacilityCompliance Act of 1992,” Section6001 of Solid Waste Disposal Act, 42USC 6961, Public Law 102-386,102nd Congress (1992).

4. “Federal Compliance with Rightto Know and Pollution PreventionRequirements,” PresidentialExecutive Order of August 3, 1993,Federal Register 58 (180), pp. 41981-41987 (August 6, 1993).

5. Board on Radioactive Waste Man-agement, “Rethinking High-LevelRadioactive Waste Disposal,” a posi-tion statement (National AcademyPress, Washington, D.C., 1990).

6. J. D. Watkins, “Posture State-ment,” DOE/CR-0011, U.S. Depart-ment of Energy, Washington, D.C.(January 1993).

7. U.S. Department of Energy, “Envi-ronmental Restoration and WasteManagement Five-Year Plan,”DOE/S-00097P, Washington, D.C.(January 1993).

7

W. Brad SmithNuclear Materials Processing:Chloride Systems

Off-Gas Scrubbing

Waste Minimization Technologies

IntroductionEffectively handling corrosive off-gases produced by aqueous-based processing of Special Nuclear Materials is an enigma for mostfacility operations. Resolving the problem is not just limited toscrubbing the corrosive off-gases. Project research and design mustalso consider containment design, ventilation control, the effects ofradiation and other contaminants, monitoring and control systems,maintenance, and waste minimization. The comprehensive plan forthe aqueous recovery operations at the Los Alamos Plutonium Fa-cility incorporates a minimum of local corrosion-resistant material,the capture of all corrosive acid vapors near the source, and moni-toring of the glove-box exhaust system. The first phase of the planconsists of consolidation of all chloride-based processing to mini-mize the number of containment systems requiring special materi-als and maintenance. The second phase involves the design andevaluation of the corrosion-resistant containment systems such asglove-box linings, dispersion-coated ventilation systems, and linedprocess piping. The third phase is comprised of various efforts toprotect the local utility support systems; these efforts include theuse of an off-gas central scrubber for the chloride recovery glove-box exhaust system and local caustic scrubbers for the chloride andnitrate vacuum transfer systems. This paper first examines the ex-tent of corrosion caused by acid vapors in the Los Alamos Pluto-nium Facility, reviews the history of integrating aqueous-basedchloride processing into our nuclear processing facility, and thenfocuses on the operation of the off-gas central scrubber, vitally im-portant to implementing the third phase of the Los Alamos planfor aqueous recovery operations.

Acid Vapor Corrosion at the Plutonium FacilityConcerns about acid vapor corrosion at the Plutonium Facility havegrown with the increase in applications for acid-based aqueousprocessing, the buildup of local corrosion, and the trend towardshorter-lived HEPA (high efficient particulate air) filters. Otherfacilities are concerned about the corrosion problems caused bythe increased use of hydrofluoric acid, hydrochloric acid (HCl),chlorine, and caustics.1 Acidic liquid corrosion may cause moreproblems than acid vapor corrosion; however, the problems causedby acid vapor corrosion can be more serious. Effects such as chlo-ride stress cracking of stainless-steel systems, HEPA filter seal ormaterial failures, and acid-level buildup in the utilities can be moredifficult to contain, monitor, or evaluate as compared to liquid-based corrosion effects.

“The comprehensive plan forthe aqueous recovery opera-tions at the Los AlamosPlutonium Facility incorpo-rates a minimum of localcorrosion-resistant material,the capture of all corrosiveacid vapors near the source,and monitoring of the glove-box exhaust system.”


Off-Gas Scrubbing


Since most standard industrytechniques for determining theextent and rate of corrosion arenot feasible in a plutonium pro-cessing facility, the Los Alamosplan for aqueous recovery op-erations is based on processknowledge and the specificsystem’s susceptibility to corro-sion. Visual inspection of theenclosures and the glove-boxexhaust plenums provides theonly evidence of the extent ofcorrosion. With the exceptionof some nonstainless appurte-nances, the glove boxes have ex-perienced the greatest corrosion.The location is typically nearthe glove-box bottom, where liq-uids collect. Similarly, the localducts are significantly more pit-ted near the duct bottoms thanthe duct tops. The scale buildupis highest nearer the glove-boxexhaust, most probably due tocondensation and the subse-quent ferric precipitates fromdissolved iron components.The rate of corrosion is also ex-tremely difficult to evaluatedue to the variations in operat-ing times, concentrations, andchemical interactions. Specificupsets with processes andequipment, such as the evapora-tion process for nitrate-basedoperations, dissolution pro-cesses, loss of condenser coolingwater, or liquid carryover fromvacuum transfer operations,may cause more corrosion thanthe continuous release of low-concentration fume gas. Cur-rently, no system or procedureis in place to monitor theseupset conditions.

The selective application of cor-rosion-resistant material in thesecond phase of the plan hasbeen technically and economi-cally feasible primarily becauseof the first-phase consolidationof aqueous chloride operations,which routinely vent low, butsignificant, concentrations ofHCl vapors. The PlutoniumFacility was originally designedfor nitric acid (HNO3) opera-tions, so even these low concen-trations pose a long-termproblem for the Facility. Corro-sion problems do exist withother operations, but the impactfrom chloride corrosion presentsthe biggest potential for failuredue to stress-corrosion cracking.Although the Plutonium Facilityhas not experienced any pri-mary containment failures be-cause of corrosion, many factorssuch as general process knowl-edge, scale buildup in the localducts, and localized glove-boxcorrosion indicate material in-compatibilities exist. Further-more, because of the variety ofprocess operations and the sizeof the Facility, replacing all ex-posed systems with special cor-rosion-resistant materials wouldnot be economically feasibleand, thus, preventative mea-sures become necessary. Preven-tative measures at Los Alamosconsist of off-gas scrubbing atthe various sources of acid va-por combined with selective ma-terial upgrades and real-time,in-line monitoring to ensure theperformance of these scrubbers.

History of Integrating Aqueous-Based Chloride ProcessingThe Los Alamos graded ap-proach to implementing anddemonstrating an integratedaqueous chloride flow sheet wasgoverned by the severity of theeffects from HCl corrosion. Thisgraded approach proceeded asfollows: early emphasis on newliquid-processing equipmentand transfer systems, followedby secondary containment ofliquids, then ventilationsystem upgrades, and finallypoint-source off-gas scrubbing.

New liquid processing equip-ment and transfer systems weredeveloped using chlorinatedand fluoropolymer-based plas-tics, such as CPVC (chlorinatedpolyvinyl chloride) end plates,commercial-grade plastic pipingand valves for internal glove-box transfer systems, PVDF(polyvinylideneflouride) slabtanks for process storage, andplastic-lined piping for all exter-nal liquid transfer systems.This use of plastics reducedcorrosion at the most susceptiblepoints, provided a level of safetyand reliability similar to typicalmetal systems, and actually im-proved accessibility and flexibil-ity in processing.

Aqueous batch processing in aresearch and development facil-ity can result in spills of minoramounts of process liquids be-cause of process-tubing changesand routine operations or main-tenance; thus, secondary con-tainment of liquids is essential.

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Drawing on experience gainedwith fluoropolymers for processequipment, we usedfluoropolymer-based plasticsto protect enclosures, appurte-nances, and utilities. The glove-box shell, which provides thesecondary containment, has his-torically been fabricated of 316stainless steel, as are most ap-purtenances. The published cor-rosion rate for 316 stainless steelin concentrated (10 M) aqueousHCl is at least 0.05 inches peryear2 (versus a rate of 0.002inches per year for aqueousHNO3). This corrosion rateonly provides for an approxi-mate 10-year life cycle for7-gauge stainless steel. Evalua-tion and testing of several differ-ent glove-box linings led to anew specification3 for the fabri-cation and installation offluoropolymer enclosure linings,which provide both mechanicaland chemical protection.4

Ventilation equipment has beenmodified for compatibility withlow concentrations of acid va-pors. All mild steels used forHEPA fire screens and supportshave been replaced with stain-less-steel components. How-ever, recent studies5 have showna decrease in the efficiency of thethree-stage Facility HEPA filtersfor the aqueous chloride pro-cessing ventilation system.Trends toward higher DOS* pen-etration rates are evident for thefirst-stage HEPA filters but arestill an order of magnitude be-low the design failure point.

Off-Gas Scrubbing


The consolidation of the aque-ous chloride operations providesthe opportunity to collect andtreat all HCl fugitive emissionswith a central scrubber and en-hances the life cycle of the Pluto-nium Facility systems that don’tuse chloride-resistant material.

An evaluation of commercialfume gas scrubbers found onesystem that met the stringent re-quirements for integration withthe Plutonium Facility systems.Some of the features providedby this system include auto-matic, unattended operation;low maintenance; critically safeliquid waste storage; wasteminimization; and fail-safe shut-down. The system is sized totreat the exhaust from four tofifteen glove boxes and has a to-tal system pressure loss abouthalf the available static pressureof the glove-box exhaust system.Because this scrubbing processprovides a utility support func-tion, which only enhances thesafety and life cycle of the Facil-ity, it can be easily automatedwithout the burden of a “safetyclass” designation.

The HCl central scrubber isdesigned to remove acid gasfumes from twelve glove boxes.The process equipment will beinstalled in specially designedenclosures to meet ALARA (aslow as reasonably achievable)contamination control goals as-sociated with maintenance op-erations and instrumentcalibration. Fumes will be col-lected in a local corrosion-resis-tant exhaust duct, removed byliquid absorption in a packedcolumn scrubber, and thenvented to the glove-box exhaustsystem. (Refer to Fig. 1 for thisand other design details thatfollow.)

Although the exact cause hasnot been determined, the de-crease in filter efficiency corre-sponds to an increase inaqueous operations in this area.To trace possible causes, accessports for periodic sampling ofthe ventilation exhaust for acidgas, for measurement of exhaustflow rates, for corrosion coupontests, and for visual examinationof scale buildup have been in-corporated into the local glove-box exhaust ducts, where theacid gas concentrations shouldbe the highest.

Point-source off-gas scrubbingwas originally implemented toprotect the vacuum transfersystem. Fume gas was origi-nally vented directly to theglove-box atmosphere fromdissolution, feed treatment,and evaporator processes.These concentrated fumes,which condense on and etchthe glove-box windows, re-duced visibility. The humidityalso caused the local HEPA filterto swell and plug prematurely.Local scrubbers now captureapproximately 90% of all HClacid vapors and 50% of allHNO3 vapors.

HCI Central Scrubber DesignFor some processes the vacuumtransfer system cannot providethe necessary volume to collectthe acid vapors. These fugitiveemissions are vented throughthe glove-box exhaust systemand continuously diluted by ex-haust from other areas. Al-though the local scrubbersremove a majority of the fumes,some concern about the effectsof chlorides on stainless-steelsystems, neoprene seals, andHEPA materials still exists.* di (2 ethylhexyl) sebacate, which

replaces DOP (dioctyl phthalate).


Off-Gas Scrubbing


The maximum static pressurefor the glove-box exhaust is anegative 1.3-inch water column(WC). The maximum scrubbersystem pressure loss is a 0.7-inchWC. Therefore, a very sensitiveinstrumentation system isrequired to monitor and controlthe pressure loss. Initially, thepressure will be maintained byallowing the flow to decrease asthe HEPA filters and scrubberpacking plug up. As the pres-sure loss increases above apredetermined set point, thesystem will go into a bypassmode (see Flow Control Valves,designated as FV, in Fig. 1).

Fig. 1. Scrubber Piping andInstrument Diagram.

Upset or maintenance condi-tions will also force the systeminto a bypass mode. EmergencyShutdown (ESD) logic will bedeveloped and monitoringinstalled to insure proper systemresponse. Sufficient local gaugesand manual overrides will beprovided to insure fail-safeoperation. However, failure ofany component will still notaffect the glove-box exhaustsystem flow and pressurestability, minimizing the needfor redundant or “safety-class”systems.

As HCl vapor is absorbed, thepH of the scrubber solution willdecrease. Batch processingcauses the influent concentrationof HCl to vary over a widespectrum. Therefore, the scrub-ber liquid discharge rate (see pHControl Valve, AV) will beproportional to changes in theHCl concentration, which ismonitored by measuring the pH.This sidestream of scrubbersolution will be piped to a set ofslab tanks, and an equal amountof fresh water will be added (seeLevel Control Valve, LV) tomaintain an optimum pH of 2.

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Off-Gas Scrubbing


plugging of the proposed post-scrubber HEPA. The use of apost-scrubber HEPA (whichcould be installed in one of theutility support enclosures tosimplify change out and dis-posal) between the centralscrubber exhaust and the mainducts would help insure that nocontamination would spread tothe Facility exhaust ducts.♦

Cited References1. U. S. Department of EnergyHeadquarters of Nuclear Safety,“Incident Summary,” in “Opera-tions Weekly Report, No. 92-26”(October 16-22, 1992).

2. Chemical Engineers Handbook,Fifth Edition, R. Perry andC. Chilton, Eds. (McGraw Hill, Inc.,New York, 1973), p. 23-18,Table 23-3.

3. Los Alamos National Laboratory,“Technical Specification for Enclo-sure Linings,” Specification NMT8-1166-R00 (January 1990).

4. M. Dinehart, “Corrosion Resis-tant Linings for Glove-Box Enclo-sures,” in Proceedings of NACEConference, Houston, Texas,August 1991.

The water level in the scrubbersump will have redundantmonitors. One system (seeLevel Control, LC) will be ananalog signal used as an over-ride for the pH control to main-tain proper pump suction withalarm indication for high andlow levels. A second set of levelswitches (not shown), high-highand low-low, will be used toshut down the scrubber andplace the ventilation FlowControl Valves in a bypassmode. A tertiary level control,consisting of an open drain line,shall be used to prevent anypossibility of liquid depthsabove 4 inches (because ofcriticality concerns). A flow-limiting device may also beconsidered to limit the totalamount of water used on aperiodic basis.

The primary contaminationcontrol for glove boxes is thenegative pressure for the glove-box exhaust system in theaqueous chloride processingarea. Therefore, the pressurewill be monitored (not shown),and any excursions beyond setpoint will override the bypassFlow Control Valves. A redun-dant pressure switch will acti-vate an alarm and also place theventilation Flow Control Valvesin a bypass mode.

The exhaust from the scrubbermay be saturated with water;the resultant humidity will affectHCl detectors and air flowmonitors, and condensationwill increase the possibility ofcorrosion in the main ducts.Using a local chiller to cool thescrubber solution will lower thedew point as well as increasethe removal efficiency of thescrubber.

A mass spectrometer is the mostfeasible method to monitor theeffectiveness (HCl removal) ofthe scrubbers. This type ofcontinuous, on-line instrumenta-tion is required to evaluatemultiple acid-gas concentra-tions, provide wide rangeabilityto determine upset conditions,and provide detailed glove-boxexhaust air analysis on a con-tinuous basis.

Waste MinimizationTo minimize liquid waste, thecentral scrubber will have aliquid discharge rate that isproportional to changes in thesystem pH, as previously dis-cussed. This control system willreduce the projected waste from4000 gallons per month toapproximately 120 gallons permonth. This waste stream willbe neutralized and then sentthrough the caustic waste line tothe Los Alamos Liquid WasteTreatment Facility. One possiblealternative to treating the wasteis to recycle or reuse the scrub-ber solution and thereby elimi-nate liquid waste altogether.

Another approach being consid-ered for solid waste minimiza-tion is to replace local glove-boxHEPA filters, which requirefrequent changing because ofswelling due to the absorptionof moisture from the aqueousoperations, with corrosion-resistant roughing filters and apost-scrubber HEPA. Roughingfilters will prevent alpha con-tamination of the local ducts dueto particulates, provide moreconsistent flow and more consis-tent pressure control, and reduce


Acid Recovery and RecycleIntroductionPlutonium processing, both for waste handling and for recovery,uses large amounts of nitric and hydrochloric acids. Typical pro-cesses involve dissolving some form of plutonium in these acidsfor purification. These acid-based processes effectively accomplishneeded separations, but they generate large volumes of waste acidsolution.

The previous approach to handling these solutions has been toneutralize them with caustic and then partition radioactive compo-nents from the bulk of the solution. The radioactive fraction is putinto a concrete matrix (saltcrete), and the remaining solution is dis-charged to the environment, after treatment to remove radioactivity.At Los Alamos, the bulk of nitric acid (HNO3) is separated fromnonvolatile salts (including most of the plutonium), but both frac-tions are presently disposed of rather than recovered and recycled.

This approach leads to environmental and storage problems.With both our heightened commitment to protecting the environ-ment and future regulatory requirements, reduction of generatedwastes is essential. Including handling, inspection, packaging,shipping, and actual storage, the cost of producing waste units canbe tremendous. Beyond cost implications, concerns over long-termstorage of radioactive waste dictate that both the amounts be mini-mized and that the storage forms be highly stable. In particular, apotential for leaching of soluble salts from saltcrete exists, and suchleaching could render the storage form less stable.

Recycling acids from waste solutions helps solve many of theseproblems because this method eliminates most of the waste gener-ated by plutonium processing. Hazardous chemical discharges(notably nitrates) are greatly reduced, and already low levels ofradioactive discharges are further reduced. This reduction in thetotal number of radioactive waste units requiring storage lowersdisposal costs significantly; for example, acid recycle would saveover $50 million in waste disposal costs in five years for a facilityoperating at levels similar to the previous levels of the Rocky FlatsPlant. The reduced salt content in radioactive wastes, an additionalbenefit of acid recovery and recycle, reduces the potential forleaching of stored radioactive wastes.

Minimizing the concentrations and amounts of plutonium and neu-tralized salts sent to waste disposal supports the Nuclear MaterialsTechnology (NMT) Division’s vision of zero radioactive and hazard-ous discharges.

Thomas R. MillsActinide Materials Chemistry

“Minimizing theconcentrations andamounts of plutoniumand neutralized salts sentto waste disposal supportsthe Nuclear MaterialsTechnology Division'svision of zero radioactiveand hazardous discharges.”


13

Our three-year goals are1) a 90% recycle of HNO3,2) a 90% reduction of acid concentration in waste streams, and3) a 10-fold reduction of nitrate concentration in effluent liquid.

Our five-year goals are1) an operating recycle system for both acids that will eliminate effluent acid streams,2) a reduction of nitrate wastes by a factor of 500, and3) a reduction of chloride wastes by a factor of 10.

ApproachA two-stage recovery and re-cycle process is planned. First,each acid waste stream is evapo-rated to separate volatiles (pri-marily acid and water) fromdissolved radioactive solids andother inorganic salts. Second,the evaporated acid and waterare separated in a fractional dis-tillation column to yield a con-centrated acid and a waterstream. This two-stage processis shown by the block diagramin Fig. 1.

The nonvolatile salt residuefrom evaporation, which con-tains any plutonium and inor-ganic salts from solution,represents a small fraction of theoriginal acid solution. This frac-tion will be neutralized anddried and may receive furthertreatment (e.g., denitration orfreeze drying1) before disposal.The resulting solid will be dis-posed of in a concrete matrixor an alternative ceramic form.

EvaporatorDistillationColumn

Water

RecycleAcid

Waste AcidSolution

NonvolatileSalts

VolatileComponents

Fig.1. Block diagram of acid recovery and recycle system.

Acid Recovery and Recycle


The acid/water vapor mixturefrom the evaporator will beseparated in the distillation col-umn to yield a concentrated acid(≥12 M HNO3, 6 M HCl) at thecolumn bottom and a very di-lute acid stream (“water”) at thecolumn top. The separated acidwill be sufficiently concentratedto be reused for processing, andit will contain minimal pluto-nium or other constituents.The water stream will requirepH adjustment to neutralityprior to discharge to the envi-ronment. This effluent streamwill have a negligible plutoniumcontent and will have a salt con-tent much less than at present.

Nitric and hydrochloric acidrecovery and recycle will bedemonstrated at the Los AlamosPlutonium Facility using the twointegrated demonstration lines:Advanced Testing Line for Ac-tinide Separations (ATLAS) for

HNO3 and Experimental Chlo-ride Extraction Line (EXCEL)for HCl. ATLAS already has athermosyphon evaporator as anintegral part; a distillation col-umn will be added as an up-grade. HCl recycle is at an earlystage; thus, characterization ofHCl processes (which includesflow balances) and bench-scaleevaporation and distillationexperiments are needed. Bothan evaporator and a distillationcolumn will be added to EXCEL.

Following demonstrations usingthese systems, full-scale systemswill be built to handle all acidwaste streams from the Pluto-nium Facility. The acid recoveryand recycle effort will have ac-tive participation from membersof the Nitrate Systems (NMT-2),the Chloride Systems (NMT-3),and the Actinide MaterialsChemistry (NMT-6) groups.


Project StatusThe Nitrate Systems Group, us-ing the ATLAS thermosyphonevaporator, has already con-ducted valuable initial experi-ments with HNO3. Promisingresults indicate that acid recycleis feasible. A simulated processwaste solution containing 4.5 MHNO3 was evaporated with thegoal of separating the maximumpossible fraction of containedacid. At evaporation rates of0.7–0.8 L/min, over 82 mol %of the HNO3 was recovered forfurther concentration. The vol-ume of solution remaining in theevaporator was only 2% of theamount fed to the evaporator,and this solution was flushedfrom the evaporator before theacid concentration process wasbegun. In a second pass of cleandistillate through the evapora-tor, steady HNO3 concentrationswere reached within 30 minutes.In this second pass, 43 mol % ofthe original nitric acid feed wasconcentrated to 10.7 M HNO3.In a third pass, 21 mol % of theoriginal nitric acid was collectedas 12 M HNO3.

A physical design for acid re-cycle systems has been devel-oped for use in the presentATLAS system. The HNO3distillation column is closelycoupled to the ATLAS evapora-tor to take advantage of thermalenergy invested. Vapor exitingthe evaporator becomes thefeed to the column. The existingthermosyphon evaporator usesa steam bundle to producevigorous boiling and liquidcirculation through the boiler.

Demister

SteamBoiler

LiquidLevel

Waste AcidSolution

Condenser

SteamBoiler

Condenser

Water

RecycleAcid

NonvolatileSalts

Fig. 2. Schematic of thermosyphon evaporator and acid distillationcolumn.

Important to retention of radio-active components, a demisterbed in the evaporator minimizesentrainment of fine dropletswith dissolved solids.

The distillation column containsloose-poured packing to supporta dispersed liquid phase trick-ling downward. The packingprovides maximum interfacialarea for vapor-liquid equilib-rium but offers minimal resis-tance to upward vapor flow.The column has a steam boilerfor vapor boilup and a partialcondenser for liquid reflux.A schematic diagram of theevaporator and column isshown in Fig. 2. The evaporatorand column combination forEXCEL will have a similarconfiguration.

Preliminary designs have beenmade for HNO3 and HCl distilla-tion columns for ATLAS and EX-CEL. The design for a HNO3distillation system is based onhandling 1 L/min of 4–6 MHNO3. A 6-inch-diameter,6-foot-tall distillation columncan concentrate at least 95% ofthe HNO3 in this stream to 12 MHNO3. The waste stream willhave less than 5% of the acidfeed, and plutonium activity willbe lower than that obtained usingevaporation only. “Cold” testsare under way on the HNO3distillation column to determinepacking performance for flowcapacity and separating ability.Following these tests, the columnwill be installed in ATLAS.



15

(In parallel, we are discussingwith a vendor possible purchaseof acid recovery systems thatuse technology developed atPacific Northwest Laboratories.)A similar system for HCl distil-lation is expected to produce6 M HCl from a feed of 1 L/minof 2–4 M HCl.

The highest volume of theHNO3 effluent stream from thePlutonium Facility in recentyears was about 75,000 liters peryear. Although the capacity ofATLAS (~0.75 L/min) mightconceivably handle this stream,the required duty cycle of 80%(1667 hours) is unrealistic for aplutonium facility. Full-scaleacid recycle systems will bedesigned and built usingexperience gained from theATLAS and EXCEL demon-strations. These systems willaddress the five-year goals.

SummaryDevelopment of acid recoveryand recycle will enable NMT Di-vision to reduce wastes on a fa-cility-wide basis. Our researchwill demonstrate that hazardousand radioactive chemicals can bereduced and eliminated from liq-uid effluent streams in an exist-ing plutonium facility. Inaddition, this important technol-ogy will be available for the plu-tonium facility yet to come —Complex 21.✦

Cited Reference1. N. V. Coppa, “Application ofFreeze Drying Technology toDecontamination of RadioactiveLiquids,” this report.




Louis D. Schulte,Richard R. Salazar,and Benjie T. MartinezNuclear Materials Processing:Chloride Systems

John R. FitzPatrick,Bradley S. SchakeInorganic and StructuralChemistry GroupChemical Science and TechnologyDivision

Hydrochloric Acid Waste Stream Polishing

“The advantages of apply-ing chromatographic tech-niques to decontaminationof radioactive aqueouseffluents include generationof smaller quantities of solidresidues in forms suitablefor storage, reduction ofTRU waste volumes, andmore efficient decontamina-tion of aqueous effluents.”


IntroductionAqueous processing of plutonium residues in hydrochloric acid(HCl) produces waste effluent streams that require several treat-ment steps before the liquids may be discharged to the environment.Effluents from HCl processing streams are currently combined androuted to controlled hydroxide precipitation as the first step in ac-tinide decontamination. This step neutralizes the entire inventoryof process acid, coprecipitates many other metal salts with the ac-tinides, and produces a solid cake that is often too high in ameri-cium and plutonium for eventual WIPP (Waste Isolation Pilot Plant)disposal. The liquid effluent from controlled hydroxide precipita-tion (usually about 109 dpm per liter) is the largest contributer ofalpha activity for solutions sent to the TA-50 Liquid Waste TreatmentFacility. Final treatment steps at TA-50 include a ferric flocculationprocess, which produces additional barrels of transuranic (TRU)solid waste.

The purpose of this work is to evaluate extraction chromatographytechniques and materials as an alternative method to removingactinides from aqueous Hcl effluent streams. Part of the challengehas been to develop and match extraction methods to specificwaste stream requirements for best overall actinide decontamina-tion. Effluent streams may be categorized as “high-acid streams”(~6 M HCl) which originate from solvent extraction raffinate oranion exchange effluent, or “low-acid streams” (~1 M HCl) whichmay come from the filtrate of plutonium(III) oxalate precipitationor the eluate of an anion exchange “guard column” used to removeiron from the process stream. The advantages of applying chro-matographic techniques to decontamination of radioactive aqueouseffluents include generation of smaller quantities of solid residuesin forms suitable for storage, reduction of TRU waste volumes, andmore efficient decontamination of aqueous effluents.

High-Acid Effluent StreamsThe majority of our early research has concentrated on high-acidstreams for several reasons:

1) high americium content presents the greatest decontamination challenge,2) increasingly low limits on activity content of grout waste forms indicate the need for isolation of the americium in a stable form for storage or vitrification,3) removal of americium early in the processing flowsheet will reduce personnel irradiation exposure from subsequent waste handling steps,4) decontamination of the effluent would provide feed for a simple acid recycle scheme for about 90% of the HCl used in plutonium processing, and

17

5) segregation of the high-acid stream from other effluent streams simplifies some aspects of waste polishing design.

Separation of actinides from the other elements in the wastestream, mostly alkali or alkaline earth metals, appears feasibleusing present-generation technology. CMPO (n-octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide) and relatedmolecules have been examined for use in liquid-liquid extractionschemes in hydrochloric media to remove actinides.1, 2 Extractionchromatography, using similar extractants physically sorbed to asolid support, has been successfully used to concentrate actinideson an analytical scale; however, testing of these techniques forlarger-scale applications has received scant attention.3–5 Our workhas demonstrated extraction of sizable quantities of actinides fromHCl media containing alkali or alkaline earth metals and showspromise for development into a reversible chromatographic system.

Several resins coated with CMPO diluted in either DAAP (diamylamylphosphonate) or TBP (tributyl phosphate) were tested. Con-tact studies with plutonium(IV) demonstrated a distribution coeffi-cient (Kd) of ~104 in 6 M HCl, and a Kd near 1 when reduced toplutonium(III) in 1–4 M HCl (see Fig. 1). Kd is defined as actinidequantity per gram of resin ÷ actinide quantity per milliliter of solu-tion. Flow studies of plutonium(IV) have shown decontamination

factors (DF) of up to 104. DF isdefined for a particular solutionas actinide concentration beforetreatment ÷ actinide concentra-tion after treatment. Elution ofplutonium(IV) from the loadedcolumns in flow experimentshas been slower than desiredbut is expected to be improvedby altering the elution condi-tions. Loading studies withplutonium(IV) indicate that theresins should be useful for HClprocess stream decontamina-tion. The maximum plutoniumloading observed in this studywas 107 mg per gram of resin,a loading capacity approachingone-half that of a standard an-ion exchange resin (typicallyabout 1 millimole per gram ofresin or ~240 mg actinide pergram of resin).

Americium(III) removal hasproven more difficult. Contactstudies of several CMPO-bear-ing resins with americium(III)have shown dramatic variationin observed Kd, dependentupon the amount and type ofdiluent used to disperse CMPOon the bead. Contact studiesto remove americium(III) on aresin loaded with pure CMPO(no diluent) in 6 M HCl demon-strated the largest Kd (~103),but this resin was kineticallytoo slow to be useful in a flowsystem. Other resins with betterkinetic properties gave a Kd ofabout 102, high enough to re-tain significant americium(III)but with breakthrough largerthan desired in flow experi-ments. We are continuing todesign and test new extractiveresin materials tailoredto more effectively removeamericium(III) from thesehigh-acid effluent streams.



Fig. 1. Changes in Pu retention on 20:20 CMPO/DAAP resin with NaNO2and NH2OH additions.




Fig. 2. Changes in Am(III) retention on Diphonix with Fe(III) addition.

Low-Acid Effluent StreamsThe low-acid effluent streams will require a fundamentally differentapproach for decontamination because of the presence of additionalchemical interferents. Oxalic acid competes with neutral ligandslike CMPO in complexation of actinides in HCl, especially at low-acid concentrations. Iron(III) and certain other transition metalelements, present in the effluent stream from the guard column,act as interferents for CMPO-type ligands. For these reasons weare examining Diphonix,* a commercial polystyrene resin contain-ing covalently-bound gem-diphosphonic acid groups, for theseapplications. Preliminary experiments with this resin have showna very good uptake of americium from dilute HCl solutions with ameasured contact Kd of about 105 (see Fig. 2). Other workers havedemonstrated that the Diphonix resin is capable of selectively re-moving actinide(III) ions from most other elements in acidic solu-tions and with little interference from oxalate.6,7 Elution ofactinide(III) elements from the column may be accomplished with

* Diphonix resin is commercially available from EIChroM Industries, Inc., Darien, Illinois.

very strongly coordinatingdiphosphonic acid ligands(not desirable for our process)or by flooding the columnwith selected stable elementsin dilute acid.8 Our preliminarydisplacement experiments indi-cated that Am+3 should be effec-tively eluted using Al+3 or Fe+3.Further work is required to fullyevaluate the promising potentialof Diphonix resin for low-acidwaste stream treatment.

ConclusionsMajor improvements in the formand quantity of actinide wastesgenerated from HCl plutoniumprocesses are possible. Treatmentof specific process effluents atthe source, with extraction meth-ods tailored for the individualstream, offers the best chance ofsuccessful actinide decontamina-tion. The payoff will be concen-tration of actinides into smallervolumes of storable or treatablesolid forms, decontamination ofhigh-acid stream effluents to ac-tivity levels that allow facile HClrecycle, reduction in alpha activ-ity in waste solutions releasedto TA-50, and reduction in thevolume of TRU solid wastegenerated.✦

19

5. G. S. Barney, and R. G. Cowan,“Separation of Actinide Ions fromRadioactive Waste Solutions UsingExtraction Chromatography,”Westinghouse Hanford report,WHC-SA-1520-FP (1992).

6. E. P. Horwitz et al., “Uptake ofMetal Ions by a New Chelating Ion-Exchange Resin. Part 1: Acid De-pendencies of Actinide Ions,”Solvent Extr. Ion Exch. 11 (5), 943-966 (1993).

7. R. Chiarizia et al., “Uptake ofMetal Ions by a New Chelating Ion-Exchange Resin. Part 2: Acid De-pendencies of Transition Metal andPost-Transition Metal Ions,” SolventExtr. Ion Exch. 11 (5), 967-985 (1993).

8. E. P. Horwitz, Argonne NationalLaboratory, personal communica-tion, August 1993.



Cited References1. E. P. Horwitz, H. Diamond, andK. A. Martin, “The Extraction of Se-lected Actinides in the (III), (IV) and(V) Oxidation States from Hydro-chloric Acid by OFD(iB)CMPO:The TRUEX-Chloride Process,”Solvent Extr. Ion Exch. 5 (3), 447–470(1987).

2. E. P. Horwitz, H. Diamond,and K. A. Martin, “Extraction ofAmericium(III) by Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphineoxide,” Solvent Extr. Ion Exch. 5 (3),419–446 (1987).

3. W. I. Yamada, L. L. Martella, andJ. D. Navratil, “Americium Recoveryand Purification Using a CombinedAnion Exchange-Extraction Chroma-tography Process,” Journal of the LessCommon Metals 86, 211–218 (1982).

4. G. J. Lumetta, D. W. Wester,J. R. Morrey, and M. J. Wagner,“Preliminary Evaluation of Chro-matographic Techniques for theSeparation of Radionuclides fromHigh-Level Radioactive Waste,”Solvent Extr. Ion Exch. 11 (4), 663–682(1993).



A Novel Liquid-Liquid Extraction Technique

Daniel J. Kathios,Gordon D. Jarvinen,Stephen L. Yarbro,and Barbara F. SmithNuclear Materials Processing:Nitrate Systems

IntroductionThe cleanup of sites where nuclear materials were used and thetreatment of wastes accumulated during the Cold War require thedevelopment of cost-effective and efficient technologies for the pro-cessing of radioactive wastes. Such technologies separate radioac-tive materials from the nonradioactive materials that make up thebulk of the wastes. The radioactive materials are sometimes furtherpartitioned into their long-lived and short-lived radioactive compo-nents. The wastes can then be disposed of in a less expensive fash-ion and in a manner that significantly reduces their potential threatto the environment. Some of the radioactive materials may also besuitable for reuse.

A variety of liquid-liquid extraction (LLE) technologies are used forthe processing of nuclear materials. LLE is a process in which twoimmiscible liquids (commonly an aqueous and an organic phase)are allowed to come in contact so that a solute (substance dissolvedin a solvent) present in one of the phases can be transferred to theother. The overall process is composed of two separate steps —extraction and back-extraction. In the extraction step, an aqueousacidic stream, in which is dissolved a mixture of radioactive andnonradioactive species, comes in contact with an organic streamcontaining an extractant specifically designed to remove these ra-dioactive metals. In the back-extraction step, this organic streamcomes in contact with a less acidic stream into which the extractantliberates these radioactive metals. The metals from both aqueouseffluent streams (leaving the extraction and back-extraction steps)are then recovered by evaporation, precipitation, electrodeposition,or other methods.

An LLE technology best suited for the applications of interest tothis research must satisfy a variety of criteria. The technology mustoperate in a continuous mode so as to process large amounts ofwastes in a reasonable amount of time. It must be resistant to thehighly acidic and radioactive environments to which it will be ex-posed. And, because many applications will be in a contained envi-ronment, the technology must require little space, be safe andsimple to use, and require little or no maintenance. LLE technolo-gies being considered for these applications include mixer settlers,pulsed columns, spray columns, and centrifugal contactors. Eachof these technologies uses its own mode of phase mixing to expe-dite mass transfer of the solute between the two phases. 1,2 Butunfortunately, this mixing promotes aqueous/organic phase disper-sions that not only deplete the organic from the LLE process butalso reduce its overall extraction and back-extraction capabilities.

“LLE is a process in whichtwo immiscible liquids(commonly an aqueousand an organic phase) areallowed to come in contactso that a solute present inone of the phases can betransferred to the other.”

21

Fig. 1. Diagram of a porous hydrophobic membrane. The membraneprovides a region of stable and intimate contact between the aqueous andorganic phases where dispersion-free mass transfer can take place.



This dispersion problem can be avoided by using a recently devel-oped technology called a microporous hollow fiber (MHF) mem-brane module. This technology does not require phase mixing forgood mass transfer. Instead, the aqueous and organic phases comein contact with each other through a rigid, thin, and porous mem-brane as illustrated in Fig. 1. The membrane in this diagram is

made of a hydrophobic material which allows the organic phase toenter its pores or “wet” the membrane. When a slight positive pres-sure is induced on the aqueous phase, the organic phase is heldwithin the pores, and a region of stable and intimate contact be-tween the two phases is established where dispersion-free masstransfer can take place. An MHF module is illustrated in Fig. 2.It consists of an outer casing containing a large number of hollowfibers fabricated from this porous membrane material. The aqueousphase is passed through the inside of the hollow fibers (on the tubeside), and the organic phase is passed over the fiber bundle (on theshell side).

MHF modules possess a numberof features that make them anattractive technology for LLEprocesses.3-5 The small size ofthe hollow fibers permits a de-sign that achieves a very highinterfacial area in a given vol-ume, which helps to minimizethe size of the LLE process.Extractions and back-extrac-tions are conducted in a disper-sion-free fashion, thus avoidingphase coalescence problemscommonly found in other tech-nologies. There is no need tohave a difference in aqueousand organic phase densities, al-lowing for a wider array of or-ganic systems to be taken intoconsideration. The aqueous/organic phase ratios can be eas-ily varied over a wide range bymerely adjusting the flow rates.The modular nature of the unitsmakes it easy to accurately scaleup experimental results to pro-cess applications and to engi-neer any multiple moduleconfiguration that would opti-mize the process. Finally, themodules are inexpensive (capi-tal costs are low); they have nomoving parts (maintenancecosts are low); and once run-ning, they require little supervi-sion, monitoring, or externalcontrol (operational costs arelow).

The purpose of this research isto evaluate the potential for us-ing MHF modules for the pro-cessing of radioactive wastes.This is to be accomplished byanalyzing the extraction andback-extraction capabilities ofthe modules for different or-ganic systems over a range ofaqueous and organic flow rates.


Fig. 3. The complexation ofneodymium with CMPO. Threeextractant molecules are requiredto complex each neodymium ion.


The results are used to determine the number of modules requiredto achieve one theoretical extraction stage, and to provide valuableinsights as to how this technology can be improved and optimizedfor process-scale applications.

In this work, MHF modules are used to extract neodymium froman aqueous 2 M nitric acid stream and to concentrate the neodym-ium into a different aqueous stream by back-extracting it into0.01 M nitric acid. Although neodymium is not radioactive,its behavior is similar to that of trivalent americium (a long-livedactinide) and thus serves as a surrogate for this evaluation.DHDECMP (dihexyl N,N-diethylcarbamoylmethylphosphonate)

Fig. 2. Diagram of an MHF mod-ule. The design achieves a highinterfacial surface area per unitvolume and allows the aqueous andorganic phases to contact in acontinuous-flow fashion. C fi andC fo are the respective solute concen-trations in the aqueous feed andeffluent, and Cei and Ceo are therespective solute concentrations inthe organic feed and effluent.

and CMPO [n-octyl(phenyl)N,N-diisobutylcarbamoylmethylphosphineoxide] are the organic phaseextractants used in this study.These two extractants arechosen for their ability to extractneodymium at modest acid con-centrations (2 M nitric acid) andfor their high selectivity for ac-tinides over fission products.Each of these extractants removestrivalent metal ions from theaqueous phase by forming a com-plex as shown in Fig. 3. The re-verse reaction occurs during theback-extraction process. Addi-tional information about theseextractants is available in theliterature.6-9


23

˙ M = Qaq C fi − C fo( )

˙ M = Qor Ceo − Cei( )

˙ M = Kw Ai∆Clm

∆Clm =∆C1 − ∆C2

ln ∆C1 ∆C2( )



C fiinterface ≅

Ceo

m

∆C1 = C fi − C fiinterface

Theory and Definitions

General Equations

The MHF module illustrated in Fig. 2 permits a continuous-flowcountercurrent extraction (or back-extraction). The rate at which thesolute is transferred from the aqueous phase to the organic phasecan be determined by conducting a material balance around theaqueous stream. Doing so yields

, (1)

where ˙ M is the rate of mass transfer from the aqueous to the organicphase, Qaq is the aqueous flow rate, andC fi andC fo are the respec-tive solute concentrations in the aqueous feed and effluent. The rateof mass transfer can also be determined by conducting a materialbalance around the organic stream. This approach yields

, (2)

whereQor is the organic flow rate, andCei andCeo are the respectivesolute concentrations in the organic feed and effluent. An additionalway to characterize the rate of mass transfer is in terms of an overallmass transfer coefficient. For an extraction in which the aqueousphase is flowing through the inside of fibers fabricated from a hy-drophobic membrane material, this rate is expressed as

, (3)

where Kw is the overall mass transfer coefficient of the module basedon the aqueous phase, Ai is the total membrane surface area of themodule based on the internal fiber diameter, and ∆Clm is the loga-rithmic mean concentration difference given by

. (4)

∆C1 is the driving force for mass transfer at the aqueous inlet and isequal to the solute concentration in the aqueous feed minus the in-terfacial aqueous phase solute concentration at the aqueous inlet:

. (5)

Determining C fiinterface is far from trivial, and its value is commonly

approximated as the concentration that is in equilibrium with thesolute concentration in the organic effluent. This is expressed as

, (6)


where m is the distribution coefficient defined as

. (7)

Equation 5 can thus be expressed as

. (8)

∆C2 is the driving force for mass transfer at the aqueous outlet and(in a manner consistent with the determination of ∆C1 ) is expressedas

. (9)

Fig. 4. Resistances to mass transferin the extraction and back-extraction processes.



˙ M =∆Clm

1 Kw Ai( )

m =

solute concentration

in organic phase

solute concentration

in aqueous phase

equilibrium

∆C1 = C fi −Ceo

m

∆C2 = Cfo −Cei

m

A more comprehensive devel-opment of the general equationsthat characterize two-phasecountercurrent flow extractionand back-extraction processescan be found in Bennett.10

Resistances to Mass Transfer

Mass transfer of a solutethrough a porous hydrophobicmembrane occurs in three fun-damental steps. First, the solutetravels through the aqueousphase to the aqueous/organicinterface where complexationtakes place. Second, the solutecomplex diffuses through themembrane pore. Third, the sol-ute complex travels through theorganic medium. ExpressingEquation 3 as

(10)

shows that the rate of masstransfer is equal to a drivingforce term, ∆Clm , divided by aresistance term, 1 Kw Ai . Thelatter term incorporates theindividual resistances to masstransfer existing in each of thethree fundamental steps and is thus referred to as the overallresistance to mass transfer.These resistances are illustratedin Fig. 4. Equations providedby Prasad3,4 and Sirkar3-5 makeit possible to quantify theseresistances.

25

For modules fabricated with hydrophobic membrane fibers andassembled for tube side aqueous flow and shell side organic flow,these resistances can be expressed as

. (11)

In this equation, kwt is the mass transfer coefficient of the solute inthe aqueous phase inside the fibers, kos is the mass transfer coeffi-cient of the solute complex in the organic phase outside the fibers,di is the internal diameter of the fiber, do is the outer diameter of

the fiber, Dc,or is the diffusioncoefficient of the solute complexin the organic, is the mem-brane porosity, and is thetortuosity of the membranepores. Although Equation 11 ispresented here in terms of theextraction process, its applicabil-ity is equally valid for the back-extraction process as well.

Number of Modules per Theo-retical Stage

A theoretical stage in an LLEprocess is one that first allowsmass transfer between the aque-ous and the organic phases toreach equilibrium and then al-lows complete ( i.e., dispersion-free) separation of the twophases to occur before they exitthe process. A theoretical stagewith continuous aqueous andorganic flow is illustrated inFig. 5. The rate at which soluteis transferred from the aqueousto the organic phase can be de-termined by conducting a mate-rial balance around the aqueousstream. Doing so yields

, (12)

where ˙ M TS is the rate of masstransfer from the aqueous to theorganic phase in the theoreticalstage, and C fo

TS is the solute con-centration in the aqueous efflu-ent leaving the theoretical stage.Conducting a material balancearound the organic stream yields

, (13)

where CeoTS is the solute concen-

tration in the organic effluentleaving the theoretical stage.

Fig. 5. An ideal theoretical stage in a continuous-flow LLE process.C fi and Cei are the respective solute concentrations in the aqueous andorganic feed streams entering the theoretical stage, and C fo

TS and CeoTS aree

the respective solute concentrations in the aqueous and organic effluentstreams leaving the theoretical stage.



1

Kw Ai

=1

kwt Ai

+di ln do / di( )2 mDc ,or Ai

+di

domkos Ai

↑ ↑ ↑ ↑ Overall Aqueous Membrane OrganicResistance (Tube Side) Resistance (Shell Side)

Resistance Resistance

˙ M TS = Qaq C fi − C foTS( )

˙ M TS = Qor CeoTS − Cei( )


Equations 12 and 13 can be used with the distribution coefficientequation, m = Ceo

TS C foTS , to solve for C fo

TS and CeoTS as a function of

inlet concentrations and flow rates. Doing so yields

(14)

and

. (15)

For a number of modules connected in series that provide the totalmembrane surface area required for a theoretical stage,the rate at which mass transfer occurs is

, (16)

where AiTS is the total membrane surface area of these modules

based on the internal fiber diameter and ∆ClmTS is the logarithmic

mean concentration difference in a theoretical stage. The latter termis given by

, (17)

where

(18)

and

. (19)

Dividing Equation 16 by Equation 3 yields

. (20)

Substituting Equation 12 for ˙ M TS and Equation 1 for , and solvingfor Ai

TS Ai yields

. (21)

where NTS is the number of modules required for a theoreticalstage.


˙ M


C foTS =

QaqC fi + QorCei

Qaq + mQor

CeoTS =

QaqC fi + QorCei

Qaq m + Qor

˙ M TS = Kw AiTS ∆Clm

TS

∆ClmTS =

∆C1TS − ∆C2

TS

ln ∆C1TS ∆C2

TS( )

∆C1TS = Cfi −

CeoTS

m

∆C2TS = Cfo

TS −Cei

m

˙ M TS

˙ M =

AiTS

Ai

∆ClmTS

∆Clm

NTS =Ai

TS

Ai

=Cfi − C fo

TS

Cfi − C fo

∆Clm

∆ClmTS

27

Experimental Details

Preparation of Aqueous and Organic Solutions

Three different organic systems are considered in this analysis:0.25 M CMPO in DIPB, 0.81 M DHDECMP in DIPB, and 0.25 MDHDECMP in DIPB. A total of 5 L of organic solution is preparedfor each extraction (or back-extraction) experiment. Additionalinformation about these chemicals is provided in Table I.

The aqueous solution for the extraction experiments is prepared bydissolving 2.5 g of neodymium nitrate hexahydrate in 5 L of 2 Mnitric acid resulting in a neodymium concentration of 166 mg/L.The organic solution is prepared by equilibrating it against an equalvolume of 2 M nitric acid. This equilibration prevents nitric acidmass transfer from affecting the neodymium extraction process.

The aqueous solution for the back-extraction experiments is pre-pared by making 5 L of 0.01 M nitric acid. The organic solution isfirst equilibrated against an equal volume of 0.01 M nitric acid andis then added to 15 g of neodymium nitrate hexahydrate, resultingin a neodymium concentration of 996 mg/L. The organic is laterdecanted from the trace water liberated during the dissolutionprocess.

A Novel Liquid-Liduid Extraction Technique


Extractant #1:Chemical dihexyl N, N-diethylcarbamoylmethyl phosphonateAbbreviation DHDECMPSupplier Occidental Chemical Corporation

Extractant #2:Chemical n-octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxideAbbreviation CMPOSupplier M&T Chemicals, Inc.

Solvent:Chemical diisopropylbenzene, 98% (approximately 50% meta, 40% para,

and 10% ortho)Abbreviation DIPBSupplier Eastman Laboratory Chemicals

Table I. Extractants and Solvent used to Make the Different Organic Systems




Model No.: 5PCM-100Description: High Efficiency Phase Contactor

Shell Characteristics:Material polypropyleneLength 20.3 cmInternal Diameter 2.43 cmOuter Diameter 3.34 cm

Fiber Characteristics:Material polypropylene (Celgard X-10)Number of Fibers 3600Effective Length 15.9 cmInternal Diameter 240 mmOuter Diameter 300 mmEffective SurfaceArea 4300 cm

2

Effective Pore Size 0.05 µmMembrane Porosity 0.30Membrane Tortuosity 2.6

Extraction and Back-ExtractionExperiments

The experiments are conductedusing the Hoechst CelaneseLiqui-Cel Laboratory LLE Sys-tem. The MHF modules subjectto analysis are Hoechst CelaneseLiqui-Cel High Efficiency PhaseContactors. These modules aresmall laboratory-scale versionsthat are specifically designed forexperimental purposes. Theirsmall size makes it possible toevaluate the performance of thetechnology without having toprepare large amounts of aque-ous and organic feed solutions.Additional information aboutthe Hoechst Celanese Liqui-Celmodule is provided in Table II.

A schematic of the experimentalsetup is illustrated in Fig. 6. Inthe experiment, the aqueousfeed is pumped through thetube side of the MHF modulewhere it comes in contact withthe organic feed and neodym-ium mass transfer takes place.Upon exiting the module, theaqueous stream is passedthrough a flowmeter and thenthrough a flow cell present in adiode array spectrophotometerprogrammed for continuous, on-line neodymium concentrationmeasurement. The stream isthen collected in the aqueouseffluent reservoir. The organicfeed is pumped through theshell side of the module, ispassed through a flowmeter, andis ultimately collected in the or-ganic effluent reservoir. Valvesand pressure gauges are presentto control the flow rates and toensure that a positive pressureof approximately 5 psig is main-tained on the aqueous side ofthe membrane.

Table II: Characteristics of Hoechst Celanese Liqui-Cel MHFModule

During the first half of each experiment, the organic flow rate isheld constant at 5 mL/minute, and the performance of the moduleis observed over a range of aqueous flow rates. For the second half,the aqueous flow rate is held constant at 8 mL/minute, and themodule’s performance is observed over a range of organic flowrates. After a particular aqueous flow rate and organic flow rate areestablished, a period of time must elapse for the system to reachsteady state. This point is determined when the aqueous effluentneodymium concentration (measured as spectrophotometric absor-bance at 578 nm) becomes constant with respect to time. At thispoint, 5 mL of the aqueous effluent is removed from the sampleport with a syringe and is placed in a 20-mL screw-cap scintillationvial for subsequent determination of its neodymium concentration.To ensure repeatability, three 5-mL samples are taken at each steadystate.

29

Fig. 6. Experimental setup used toevaluate the extraction and back-extraction capabilities of the MHFmodules.



Evaluation of Distribution Coefficients

The distribution coefficients are determined by using smallamounts of the aqueous and organic feed solutions prepared for theextraction and back-extraction experiments. In this determination,10 mL of the aqueous solution and 10 mL of the organic solutionare placed in a 20-mL screw-cap scintillation vial and are thor-oughly mixed for one minute using a vortexer. The mixture is thenallowed to sit for two days to ensure that complete phase separa-tion has taken place. Using a syringe, 5 mL of the aqueous fractionis removed and is placed in a new screw-cap scintillation vial forsubsequent determination of its neodymium concentration. Fourdistribution coefficient determinations are conducted for each aque-ous/organic system.

Neodymium ConcentrationDetermination

Each extraction (or back-extrac-tion) experiment generates morethan 36 aqueous samples ofunknown neodymium concen-tration. Because the absorbancepeak for neodymium (exhibitedat 578 nm) is very weak, an ac-ceptably precise determinationof its concentration in thesesamples is not feasible by directspectrophotometric measure-ment. However, neodymiumdoes form a complex witharsenazo III that exhibits astrong absorbance peak at655 nm.11 Taking advantageof this, the unknown concen-trations can be accuratelydetermined.

In this approach, 0.1 g/L ofarsenazo III is added to a buffersolution of 1 M glacial aceticacid and 0.5 M sodium acetate.A 10-mL aliquot of this solutionis placed in a 20-mL screw-capscintillation vial followed by500 L of the aqueous sampleof unknown neodymium con-centration. The resulting solu-tion is thoroughly mixed forone minute using a vortexerand is allowed to sit for oneday to ensure that complexationhas reached equilibrium. Theabsorbance at 655 nm is thenmeasured with the spectropho-tometer. The neodymium con-centration of each unknown isdetermined by comparing itsmeasured absorbance with thoseobtained from a series of calibra-tion standards.


Table III. Distribution Coefficients of Neodymium for ThreeDifferent Organic Systems



Organic Distribution DistributionSystem Coefficient for Coefficient for

Extraction Back-Extraction(2 M HNO

3) (0.01 M HNO

3)

0.25 M CMPO 8.8 3.2 x 10-2

in DIPB

0.81 M DHDECMP 3.6 1.3 x 10-3

in DIPB

0.25 M DHDECMP 0.25 5.8 x 10-5

in DIPB

Results and DiscussionAll extractions and back-extractions conducted with the MHFmodules were clean, easy to run, and absolutely dispersion-free.No measurable loss of organic was detected during any of theexperiments.

The distribution coefficients for the different organic systems areshown in Table III. A superior organic system for extraction is onein which the solute preferentially dissolves in the organic phaseand thus exhibits a large distribution coefficient. From the organicsystems considered, the best for this purpose is the 0.25 M CMPOin DIPB, next is the 0.81 M DHDECMP in DIPB, and last is the0.25 M DHDECMP in DIPB. An organic system best suited forback-extraction is one in which the solute preferentially dissolves inthe aqueous phase and thus exhibits a small distribution coefficient.The best system for this purpose is the 0.25 M DHDECMP in DIPB,next is the 0.81 M DHDECMP in DIPB, and last is the 0.25 M CMPOin DIPB. Under the conditions used, the organic system with thegreatest affinity to remove neodymium during the extraction pro-cess is also the one least inclined to liberate the neodymium duringthe back-extraction process.

31

Fig. 7. Extraction of neodymium from 2 M nitric acid. Neodymiumconcentration in aqueous effluent is shown as a function of aqueous flowrate (left) and as a function of organic flow rate (right) for the differentorganic systems. Aqueous feed concentration is 166 mg/L.

The results of the extraction and back-extraction experiments areillustrated in Figs. 7 and 8. As shown in the graphs, low aqueousflow rates and high organic flow rates are best for extractions,whereas low organic flow rates and high aqueous flow rates arebest for back-extractions. The low flow rates of the solute-ladenstreams (those from which the neodymium is to be removed) pro-vide the high residence times necessary for substantial mass trans-fer to take place. In addition, the high flow rates of the soluteremoval streams (those that enter the process with no solute) main-tain low concentrations of solute in those streams and thus maxi-mize the concentration gradients for high mass transfer rates.

Figure 7 shows that an organic system with a larger distribution co-efficient provides for a better extraction. This behavior can be ex-plained by analyzing the individual resistances to the mass transferprocess. As shown in Equation 11, the membrane resistance andthe shell side resistance are both smaller for an organic system hav-ing a larger value of m . In addition, as shown in Equations 4through 9, such a system also has a greater value of ∆Clm – thedriving force for the mass transfer process. These two factors (a de-creased resistance and an increased driving force) increase the masstransfer rate and thus result in a superior extraction process.




Fig. 8. Back-extraction of neodymium into 0.01 M nitric acid. Neodym-ium concentration in organic effluent is shown as a function of aqueousflow rate (left) and as a function of organic flow rate (right) for the differentorganic systems. Organic feed concentration is 996 mg/L.

As shown in Fig. 8, superior back-extractions are achieved with anorganic system exhibiting a smaller distribution coefficient. Equa-tion 11 shows that such a system does possess higher membraneand shell side resistances. However, the value of ∆Clm (which is in-herently negative for back-extraction processes) is more negativefor smaller values of m and thus provides a stronger driving forcefor the back-extraction process. For the systems considered, theincreased magnitude of this driving force is much greater than theincreased resistance. As a result, the systems with smaller distribu-tion coefficients exhibit higher mass transfer rates for the back-extraction processes.



33

Fig. 9. Extraction of neodymium from 2 M nitric acid. The number of modules required for one theoretical stageis shown as a function of aqueous flow rate (left) and as a function of organic flow rate (right) for the differentorganic systems.

Experimental data are used withEquation 21 to determine thenumber of modules required fora theoretical stage. As shown inFigs. 9 and 10, this number in-creases with both aqueous andorganic flow rates. Thus, longeraqueous and organic residencetimes are conducive to fewermodules.

For the two stronger extractionsystems (the 0.25 M CMPO inDIPB and the 0.81 M DHDECMPin DIPB), between one and fivemodules are required for a theo-retical stage. For the weakersystem (the 0.25 M DHDECMPin DIPB), between five andtwenty-five modules are re-quired. This provides anotherreason for using a strong organic

system for extraction. Not only does such a system remove moresolute in a theoretical stage, but also fewer modules are required forthat theoretical stage to be achieved.

Comparing Figs. 9 and 10 shows that for a given organic system,the number of modules required for back-extraction is much greaterthan that required for extraction. This large difference is due to theincreased membrane resistance existing in the back-extraction pro-cess. From Equation 11, the ratio of these two resistances can beexpressed as

. (22)

This ratio is between 200 and 4500 for the organic systems in thisanalysis. Suggestions on how this membrane resistance (and there-fore the number of modules) can be reduced are provided in ourconclusions.



membrane resistance

for back - extraction

membrane resistance

for extraction

=mextraction

mback-extraction


The modules used in this analy-sis are small laboratory-scaleversions specifically designedfor experimental purposes.In an actual process application,pilot-scale (or larger) modulescan be used, each of which pro-vides as much as 10 times themembrane surface area of alaboratory-scale version.Additional surface area can alsobe achieved by connecting anumber of modules in series.These approaches would makeit possible to conduct reasonableseparations at the high flowrates normally found in processapplications.

Fig. 10. Back-extraction of neodymium from 0.01 M nitric acid. The number of modules required for one theoreticalstage is shown as a function of aqueous flow rate (left) and as a function of organic flow rate (right) for the differentorganic systems.

ConclusionsMHF modules provide a cost-effective and convenient meansby which to conduct metal sepa-rations. But initial inspection ofthe experimental data seems toindicate that reasonable separa-tions can be achieved only atflow rates that are much lowerthan those found in conven-tional, process-scale metal sepa-ration applications. But bymodifying the number or typeof modules, the organic extrac-tion systems, or the membranematerial, this technology canalso perform well at the higherflow rates typically found inprocess-scale applications.

In addition, the modules for thisanalysis use an early version of afiber bundle design that is notconducive to good shell sideflow. According to Seibert,12this poor flow causes a signifi-cant portion of the fibers to bebypassed by the organic andkeeps the surface area availablefor mass transfer from beingfully utilized. Newer modulesare now being produced with anadvanced fiber mesh that allowsthe organic to flow more uni-formly throughout the fiberbundle. A new shell side baffletechnology has also been re-cently introduced that may sig-nificantly improve the module’sperformance.



35

The technology may also be im-proved by modifying the or-ganic system. For extraction,this could be achieved by usinga more powerful extractant orby using a higher concentrationof the extractant in the organicphase. Such an approach wouldreduce the membrane resistanceand would increase the concen-tration driving force for themass transfer process. But thereis a drawback to this approach.As mentioned in the previoussection, an organic system that ismore inclined to capture metalsduring the extraction processmay be less inclined to liberatethose metals during the back-extraction process. Comparisonof Figs. 7 and 8 shows that thisbehavior is observed in our ex-perimental results. Because aprocess-scale application of thistechnology would undoubtedlyencompass both extraction andback-extraction, the perfor-mance of the organic system inboth processes must be takeninto consideration when choos-ing an improved organic system.

required for a theoretical stagein the back-extraction process.As mentioned in the previoussection, this comparably largenumber of modules is due tothe small distribution coefficient(and thus large membrane resis-tance) found in the back-extrac-tion process. Using a hydro-philic membrane would makethe membrane resistance inde-pendent of the distribution coef-ficient and would significantlyreduce the size of the process.Flat sheet hydrophilic mem-branes fabricated frompolybenzimidazole have dem-onstrated good extraction capa-bilities for neodymium and cop-per13 and are reputed to havesignificant acid and radiationstability. MHF modules madefrom this polymer may thus bepotentially well-suited for theprocess applications of interestto nuclear facilities.✦



The performance of the modulesmay also be improved by chang-ing the membrane from a hydro-phobic to a hydrophilic material.In a hydrophilic membrane, theaqueous phase fills the pores,and the neodymium passesthrough the membrane in itsionic form. In a hydrophobicmembrane (like the polypropy-lene membrane used in thisanalysis), the organic phase fillsthe pores, and the neodymiumpasses through the membranein its complexed form (seeFig. 3). The diffusivity of neo-dymium through an aqueousphase is significantly greaterthan that of an inherently largeand sterically hindered neodym-ium complex through a viscousorganic phase. The greaterdiffusivity means less mem-brane resistance and results ina greater mass transfer rate.Changing the membrane mate-rial may also reduce the largenumber of modules presently


Carbamoylmethylphosphonates -The Effects of Size and Structureof the Extractant Molecule,”Radiochimica Acta 29, 143-147(1981).

8. E. P. Horwitz, H. Diamond,and D. G. Kalina,“CarbamoylmethylphosphorylDerivatives as ActinideExtractants,” in Plutonium Chemis-try, ACS Symposium Series,William T. Carnall and Gregory R.Choppin, Eds. (American ChemicalSociety, Washington, D. C., 1983),Vol. 216, pp. 433-450.

9. E. P. Horwitz, D. G. Kalina, L.Kaplan, G. W. Mason, and H.Diamond, “SelectedAlkyl(phenyl)-N,N-dailkylcarbamoylmethylphosphineOxides as Extractants for Am(III)from Nitric Acid Media,” Sep. Sci.Technol. 17 (10), 1261-1279 (1982).

10. C. O. Bennett and J. E. Meyers,Momentum, Heat, and Mass Transfer,3rd ed. (McGraw-Hill, Inc, NewYork, 1982) pp. 598-623.

11. S. B. Savvin, “Analytical Use ofArsenazo III: Determination ofThorium, Zirconium, Uranium, andRare Earth Elements,” Talanta 8,673-685 (1961).

12. A. F. Seibert, X. Py, M. Mshewa,and J. R. Fair, “Hydraulics andMass Transfer Efficiency of aCommercial-Scale MembraneExtractor,” Sep. Sci. Technol. 28(1-3), 343-359 (1993).

13. D. Y. Takigawa, “The Effect ofPorous Support Composition andOperating Parameters on thePerformance of Supported LiquidMembranes,” Los AlamosNational Laboratory reportLA-12027-MS UC-704 (February1991).


Cited References1. Handbook of Solvent Extraction,T. C. Lo, M. H. I. Baird, andC. Hanson, Eds. (John Wiley &Sons, New York, 1983), pp. 275-507.

2. R. A. Leonard, D. G. Wygmans,M. J. McElwee, M. O. Wasserman,and G. F. Vandergift, “The Centrifu-gal Contactor as a Concentrator inSolvent Extraction Processes,” Sep.Sci. Technol. 28 (1-3), 177-200 (1993).

3. R. Prasad and K. K. Sirkar,“Hollow Fiber Solvent Extraction:Performances and Design,” J. Membrane Sci. 50, 153-175 (1990).

4. R. Prasad and K. K. Sirkar,“Dispersion-Free Solvent Extractionwith Microporous Hollow FiberModules,” AIChE J. 34, 177-188(1988).

5. Membrane Handbook, W. S. W. Hoand K. K. Sirkar, Eds. (VanNostrand Reinhold, New York,1992), pp. 727-763.

6. S. F. Marsh and S. L. Yarbro,“Comparative Evaluation ofDHDECMP and CMPO asExtractants for RecoveringActinides from Nitric Acid WasteStreams,” Los Alamos NationalLaboratory report LA-11191 UC-10(February 1988).

7. R. R. Shoun and W. J. McDowell,“Extraction of Trivalent Actinidesand Lanthanides by


37


Electrochemical Treatment of Mixed Hazardous Waste

Wayne H. Smith,Christine Zawodzinski,Kenneth R. Martinez,and Aaron L. SwansonNuclear Materials Processing:Nitrate Systems

IntroductionStrong oxidizing agents such as Ag(II) or Co(III) ion can be gener-ated electrochemically in nitric acid (HNO3) at the anode of a di-vided cell. These species are capable of oxidizing insoluble actinideoxides to the more soluble higher oxidation state and are beinginvestigated as a method to recover actinides from various scrapmatrices.1 The electrogenerated oxidants are also capable of oxidiz-ing various organic molecules to carbon dioxide and water and thuscan serve as a potential alternative to the treatment of mixed wastesby incineration.2-5

In the treatment process, the substrate of interest is added directlyto the anolyte where the catalyst is generated at a fixed rate by con-stant current electrolysis. The Ag(II) or Co(III) ion that is formedoxidizes the organic substrate and is itself reduced back to Ag(I) orCo(II). These ions migrate back to the anode surface where the oxi-dizing agents are regenerated and the cycle is repeated. Once theorganic component of the waste is destroyed, the radioactive com-ponent may be recovered by standard actinide recovery techniques.

Much is known about the electrochemical behavior of many organ-ics in broad areas such as electrosynthesis, reaction mechanisms,and kinetics. Extensive research has focused on the direct electro-chemical oxidation of organics, especially methanol oxidation be-cause of its importance in fuel-cell technology.6 The most basicconclusion of direct oxidation research is that each class of organicsreacts differently to a given set of oxidation conditions; indeed,some types resist oxidation altogether. 7

These unique reactions to oxidation have led us to study the medi-ated electrochemical oxidation (MEO) of several classes of organics,beginning with isopropanol. Isopropanol is a commonly used or-ganic solvent and has some probability of appearing in a mixedwaste. It also has only three carbon atoms, which limits the numberof possible intermediates that may be formed in the course of oxida-tion to carbon dioxide. This molecular structure should make iteasier to measure the mechanistic pathway and produce some in-sight into the overall oxidation process of alcohols in general. How-ever, the complete conversion of isopropanol to carbon dioxide andwater involves the transfer of 18 electrons :

(CH3)2CHOH - 18 e- + 5 H2O —> 3 CO2 + 18 H+ . (1)

Thus, considerable complexity still characterizes the oxidation of thisrelatively “simple” molecule.

“The most basic conclusionof direct oxidation researchis that each class of organicsreacts differently to a givenset of oxidation conditions;indeed, some types resistoxidation altogether.”


MEO is a two-step process, theheterogeneous electrochemicalgeneration of the catalyst at theelectrode surface and the homo-geneous reaction between thecatalyst and the organic sub-strate. In order to obtain aclearer understanding of theoverall reaction, we have at-tempted to study these reactioncomponents individually.Specifically, we investigatedthe effect of several criticalreaction parameters, such ascurrent density, anolyte tem-perature, acid concentration,and cell separator type, on theefficiency of catalyst generationin the absence of an organicsubstrate. We then repeatedthose experiments with theisopropanol present, looking atthe effect of these variables onthe catalyzed destruction of theorganic compound. Finally, weinvestigated the direct electro-chemical oxidation ofisopropanol in the absenceof an electron transfer catalyst.

Catalyst GenerationThe first step in the catalyticoxidation of organics is electro-chemical generation of the cata-lyst at the anode of a dividedelectrochemical cell. Typically,the kinetics and mechanism ofthese oxidation processes wouldbe studied using a small areaelectrode and modern electro-chemical techniques, such ascyclic voltammetry, rotatingring-disc voltammetry, etc.

As seen in Figs. 1 and 2, theefficiency for generation of bothAg(II) and Co(III) decreases withincreasing current density.The oxidation potential of Ag(I)is slightly less positive than thesolvent itself, whereas the oxida-tion potential of Co(II) is slightlymore positive than the solvent.Therefore, the generation ofCo(III) is impossible withoutsimultaneous oxidation of thesolvent, whereas there is a smallpotential range in which Ag(II)may be generated without oxi-dation of the solvent. In eithercase, increasing the currentdensity does result in increasingthe fraction of the current goingtoward the solvent oxidationprocess, decreasing the overallcurrent efficiency of catalystgeneration.

The effect of temperature on cur-rent efficiency was determinedby performing the electrolysis ata fixed current density but atthree separate temperatures: 20,35, and 50 °C. As the tempera-ture was increased, the genera-tion of both Ag(II) and Co(III)decreased, although this effectwas more pronounced forAg(II). Both catalysts have lim-ited lifetimes in aqueous solu-tion because of their ability tooxidize water with subsequentreduction back to the originaloxidation state. As the tempera-ture is increased, these parasiticreactions proceed at a faster rateso that the steady state concen-tration of the catalyst appears todecrease, even though the rateof generation remains essentiallyconstant.



However, utilizing these tech-niques usually requires a fairlyclean system in which the elec-tron transfer process of interestis far removed from other redoxprocesses in solution. Unfortu-nately, both the Ag(I) to Ag(II)and Co(II) to Co(III) oxidationsoccur at the anodic solvent oxi-dation limit so that neither givesa well-resolved voltammetricresponse.

To study the efficiency of cata-lyst generation, we then had touse a combined electrochem-istry-spectroscopy method inwhich a constant current densityis applied to the anode to carryout the oxidation reaction, andthe amount of the catalyst gener-ated is monitored by measuringthe amount of light absorbed atthe wavelength maximum of theoxidized species. Measurementof the catalyst concentration wasaccomplished by using a flow-through cell in the spectropho-tometer and pumping thesolution from the electro-chemical cell.

All of the measurements madeon the two catalyst systems werecarried out in 6 M HNO3.When we attempted to do theelectrolyses at lower solutionacidities, the catalysts formedinsoluble oxide salts at the elec-trode surface and fell to the bot-tom of the reaction cell.

39

Fig. 2. Effect of current density on Co(III) generation in 6 M HNO3.

Fig. 1. Effect of current density on Ag(II) generation in 6 M HNO3.

Carrying out this reaction in adivided electrochemical cell isnecessary to prevent the mate-rial that is oxidized at the anodefrom being re-reduced at thecathode. Typically, a ceramic frit,glass frit, or ion exchange mem-brane is used to separate the twocell compartments. We havefound that a ceramic frit separa-tor gives a higher catalyst yieldthan either a Nafion 117 orNafion 324 cation exchangemembrane.

The most likely explanation forthis observation is charge bal-ance within the cell. During elec-trolysis, positive charge buildsup in the anolyte while negativecharge accumulates in thecatholyte. With a ceramic frit,this charge imbalance is allevi-ated by the movement of posi-tively charged ions from theanolyte to catholyte and nega-tively charged ions fromcatholyte to anolyte. All ions insolution will cross the barrier atrates proportional to their ionicmobility and concentration. Pro-tons that are highly mobile andin high concentration will carrythe greatest fraction of the cur-rent in moving to the catholyte,while the remainder will be car-ried by the movement of nitrateions to the anolyte. The move-ment of catalyst ions across themembrane will be minimal.




Fig. 3. Electrooxidation of isopropanol at 500 mA/cm2.

With a cation exchange mem-brane, alleviation of charge im-balance is restricted to themovement of cations. Therefore,all of the current will be carriedby cation migration fromanolyte to catholyte. Whereasprotons will continue to be theprimary charge carrier, an in-creasing fraction of the currentwill be carried by catalyst ionmigration. This process isreadily observed when cobaltis used as the catalyst. With a fritin place as a separator, no cobaltenters the catholyte. When a cat-ion exchange membrane is used,the catholyte quickly becomespink, the characteristic color ofCo(II) in HNO3 solution.

To summarize, the efficiencies ofboth Ag(II) and Co(III) genera-tion are improved by use of lowcurrent density, low tempera-ture, a ceramic frit rather than a cation exchange membraneseparator, and in highly acidicmedia.

Isopropanol OxidationElectrochemical oxidation ofisopropanol in a well-stirred so-lution was carried out at fixedcurrent density at a platinum an-ode. Products and intermediatesof the reaction were monitoredby extracting a small sample ofthe anolyte during the run andinjecting this sample directlyinto a gas chromatograph (GC).

Because HNO3 is itself an oxi-dizer, we first determined if sol-vent oxidation of the organicsubstrate was contributing to theelectrochemical oxidation pro-cess. A small amount ofisopropanol was mixed with6 M HNO3, and the solutionwas stirred for several hours.

Samples for GC analysis weretaken every half-hour from thissolution. The fact that there wasno change in isopropanol con-centration and no evidence ofother substances forming duringthe time period suggested thatisopropanol is not attacked byHNO3.

Under an applied constant cur-rent with no catalyst present,isopropanol was oxidized veryquickly and efficiently to aceticacid (Fig. 3). The GC analysisclearly showed an initial

buildup of acetone, the two-electron oxidation product, fol-lowed by a decrease in acetoneconcentration when acetic acid,the four-electron oxidationproduct, began to appear.Evidence also existed for theformation of formic acid andformaldehyde in the reactionmixture. However, the reactionappeared to stop with the for-mation of acetic acid, whichoccurred when approximately10 equivalents of electrons permole of isopropanol had beenremoved from the solution.



41

This number is consistent with the partial oxidation of isopropanolto one equivalent of acetic acid and one of carbon dioxide, as per thefollowing equation:

(CH3)2CHOH - 10 e- + 3 H2O —> CH3COOH + CO2 + 10 H+. (2)

To further verify this result, we conducted a separate experimentstarting with acetic acid. After passing 10 equivalents of electronsper mole of acetic acid, we detected no change from the originalconcentration.

These results are explained in the reaction mechanism that follows:

The ease of oxidation of organic compounds by class appears to bealcohols > ketones, aldehydes > carboxylic acids. Isopropanol isvery rapidly oxidized to acetone and more slowly to acetic acid.But a product of the second oxidation step is an equivalent ofmethanol. Because methanol is even more easily oxidized thanisopropanol, it is immediately converted to formaldehyde, formicacid, and ultimately carbon dioxide. Thus, only a small quantityof these substances is present in the reaction cell at any time during

the reaction. If there were anappreciable concentration ofmethanol, the acid-catalyzedesterification of acetic acidwould be expected. The prod-uct, methyl acetate, does notappear in the chromatogram.




Conversion of isopropanol toacetone and/or acetic acid oc-curs more efficiently at highercurrent density (Figs. 4 and 5).This higher efficiency may bedue to (1) the higher acquiredpotential of the working elec-trode, which could more effec-tively oxidize the acetoneintermediate, or (2) the genera-tion of hydroxyl radicals by di-rect water oxidation, which maycontribute to the overall oxida-tion mechanism.

Temperature has a significantinfluence on the oxidation rate.At room temperature some frac-tion of the acetone still remainsin the reaction mixture even af-ter the passage of 18 equivalentsof electrons. At 50 °C the acetoneis completely oxidized to aceticacid well before 18 electronshave been added. More elevatedtemperatures may be necessaryto achieve further oxidation ofthe acetic acid to carbon dioxide.These experiments are currentlyunderway.

Finally, the oxidations were re-peated using optimal conditions(high current density, high tem-perature, and high acidity) inthe presence of a catalyst.We found that the addition ofCo(III) improved the oxidationefficiency, whereas the additionof Ag(II) led to a decrease. Themost likely explanation for thisphenomena can be attributed tothe relative redox potentials ofthe systems: Ag(II) < solvent <Co(III).

Fig. 4. Effect of current density on oxidation of isopropanol to acetone.

Fig. 5. Effect of current density on oxidation of isopropanol to acetic acid.



43

The results of our investigationssuggest that the optimum condi-tions for electrochemical oxida-tion of isopropanol are highcurrent density, elevated tem-perature, high solution acidity,and Co(III) as an electron trans-fer catalyst. However, even un-der optimum conditions,oxidation to carbon dioxide isincomplete. We are still investi-gating those conditions neces-sary to achieve completeconversion.✦

Cited References

Waste Minimization Technolgies


1. E. J. Wheelwright et al.,“Kilogram-scale Demonstration of Plutonium Recovery from PlantResidues Using CEPOD II Dis-solver, Anion Exchange, OxalatePrecipitation, and Calcination toOxide,” Pacific Northwest Labora-tory report PNL-7653 UC-721(April 1991).

2. D. F. Steele, “ElectrochemicalDestruction of Toxic OrganicIndustrial Waste,” Platinum MetalsReview 34 (1), 10-14 (1990).

3. P. Cox and D. Pletcher,“Electrosynthesis at Oxide CoatedElectrodes Part 2. The Oxidationof Alcohols and Amines at SpinelAnodes in Aqueous Base,” J. Appl.Electrochem. 21, 11-13 (1991).

4. J. Farmer et al., “ElectrochemicalTreatment of Mixed and HazardousWastes: Oxidation of EthyleneGlycol and Benzene by Silver(II),”J. Electrochem. Soc. 139 (3), 654-661(1992).

5. J. C. Farmer, F. T. Wang,P. R. Lewis, and L. J. Summers,“Electrochemical Treatment ofMixed and Hazardous Waste:Oxidation of Ethylene Glycol byCobalt(III) and Iron(II),” in Electro-chemical Engineering and the Environ-ment ‘92 (Institution of ChemicalEngineers, Warwickshire, UnitedKingdom, 1992) ICHEME Sympo-sium Series No. 127, 203-214.

6. B. Beden, J. M. Léger, andC. Lamy, “Electrocatalytic Oxida-tion of Oxygenated AliphaticOrganic Compounds at NobleMetal Electrodes,” in ModernAspects of Electrochemistry, J. O’MBockris and L. Eberson, Eds.(Plenum Press, New York, 1992)Vol. 22, pp. 97-101, 234-241.

7. Organic Electrochemistry, M. M.Baiser, Ed. (Marcel Dekker, Inc.,New York, 1973) pp. 470-494.



Nicholas V. CoppaActinide Materials Chemistry

Application of Freeze Drying Technology to Decontaminationof Radioactive Liquids

Introduction and BackgroundFreeze drying is the removal of solvent or volatile components bysublimation from a frozen solution, suspension, or multicomponentmaterial. The freeze drying process, or lyophilization when it refersto biological materials, includes two independent steps: the freezingof the material and the removal by sublimation (drying) of the vola-tile component. The removal of a solvent from a material underthese conditions often results in the preservation of the solid portionof the material. This preservation aspect of freeze drying technology(FDT) has led to diverse applications that have resulted in the pro-cessing of tons of materials daily throughout the world. The objec-tive of this work is to demonstrate the applicability of FDT to thedecontamination of complex radioactive materials including liquids,slurries, and sludges containing a wide variety of constituents inwhich the decontamination levels are in excess of 100 million. WhenFDT is applied to the decontamination of multicomponent radioac-tive materials, the product is the condensate (as opposed to the solidportion of the material). This is the first application of FDT whereprimary importance is given to the condensate, and this new twisthas drawn the attention of several commercial manufacturers anddevelopers to this work. Our focus in applying this technology isthe elimination of radioactive discharges from the Los AlamosPlutonium Facility; however, successful demonstration here willmost likely lead to nuclear industry-wide applications.

Separation of an aqueous nonradioactive solvent and a radioactivesolid (solute) during the freeze drying process occurs in the follow-ing manner (please refer to Fig. 1 throughout the following discus-sion). The solution is frozen and situated in the proximity of acondenser surface. The radioactive material in the left chamber ofthe freeze drying apparatus is held at a higher temperature than thecondenser in the right chamber (T1 > T2). A concentration gradientdevelops between the two chambers due to the difference betweenthe temperature-dependent vapor pressures (p1 and p2) of the solventat the two locations. This concentration gradient drives the transportof the nonradioactive solvent from the sample chamber to the con-denser. Heat is supplied to the frozen material to allow sublimationto occur. Because the vapor pressures can be quite low, a vacuum isapplied to assist the diffusion of the sublimed solvent. Solvent re-moval is fully achieved by supplying enough heat to the frozen ma-terial to complete the sublimation process while removing enoughheat at the condenser to maintain the concentration gradient.1

“Our focus in applyingthis technology is theelimination of radioactivedischarges from theLos Alamos PlutoniumFacility; however, successfuldemonstration here willmost likely lead tonuclear industry-wideapplications.”

45


Application of Freeze Drying Technology to Decontamination of Radioactive Liquids

frozen solution temperature,surface area and concentration,and the geometry of the appara-tus. Nuclear waste streams havea relatively low solute concen-tration in comparison to, forexample, foods. Consequently,the greatest impediment to thefreeze drying rate, for nuclearwaste solutions, is the thermalconductivity of ice. This funda-mental limitation is circum-vented through engineeringsolutions, such as maximizingthe surface area, minimizingthe thickness of the sample, and

supplying radiant energydirectly to the surface of thefrozen material.

Freeze dryers are extremelyversatile and broad in theirapplications. Laboratory freezedryers are usually designed tosublime kilograms of solventper day, whereas industrialfreeze dryers are capable ofdrying thousands of kilogramsper day.4 The factor that mostgreatly influences the cost of afreeze dryer is the performancerequired by the particularapplication.

The drying rate can be deter-mined through the simulta-neous solution of the Fourierheat law, the Knudsen equation,and the appropriate conduc-tance relationship for the geom-etry of the apparatus.2,3 (SeeReference 2 for an in-depthdiscussion of transport phe-nomena, including the Fourier,Fick, Dufour, and Sorret rela-tionships.) Such analysis indi-cates that this rate is stronglydependent on the temperaturedifference between the frozensolution and the condenser, the

Fig. 1. Schematic of a freeze drying apparatus. The sample chambercontains the frozen radioactive material (speckled) and the condenserchamber houses the cold condenser surface onto which the purified ice(gray) condenses. The temperature of the frozen material (T1) is main-tained higher than that of the condenser (T2). The vapor pressure of ice atthe two locations (p1 and p2) is dependent on these temperatures, and theresultant concentration gradient causes mass transport. Supplying theheat of sublimation at the sample and removing it at the condenser main-tains the transport conditions.


Other decontamination methodssuch as ion exchange, precipita-tion, or chelation have high ionspecificity but require multiplestages, regeneration, neutraliza-tion, and/or filtration. Theyinvolve the addition of othercomponents (resins, chelatingagents, precipitating agents, etc.)that also require remediation.Integration of FDT into theseprocessing lines may greatly in-crease process efficiency anddecontamination effectiveness.

Distillation and/or evaporationdecontamination processes in-volve continuous, often turbu-lent, flow; thus, entrainmentof the radioactive componentis an unavoidable consequence.Decontamination factors of tenthousand are theoreticallyachievable for distillation, butvalues of about one to threehundred are typical in practice.The freeze drying process differsfrom distillation in that thetransport process during freezedrying involves molecular flowand the thermal energy of theradioactive material is muchlower. These features are re-sponsible for the virtual elimina-tion of solute entrainment;consequently, decontaminationfactors obtained using FDT areextremely high.

FDT can be considered safe andcompact. No high temperaturesor pressures are used. Further-more, because the freeze dryingprocess occurs in a vacuum, thefailure of a component wouldlead to an inward leak, whichreduces the potential for con-tamination outside the system.



FDT is versatile in that its appli-cation is not limited to aqueoussolutions. Organic solvents canbe separated from other liquidsor solids. Liquids processed inthis way may be reused or dis-carded as nonradioactive.

FDT drastically reduces thevolume of the waste since thevolume of solute or solid compo-nent is smaller than the solutionor wet volume. Volume reduc-tion factors greater than onethousand have been achieved inaqueous nitrate/nitric acid solu-tions (pH < 0.5),14 but the exactvolume reduction of nuclearwaste depends on the particularmoisture content. Volumes mayalso be reduced through theelimination of neutralizationsteps in transuranic (TRU)processing. Processing of pluto-nium- and uranium-contami-nated scrap often involves thedissolution of the scrap usingnitric acid. These solutions areusually neutralized using so-dium hydroxide for compatibil-ity with subsequent processsteps. A significant percentageof the wastes stored at Hanfordwas a result of chemical neutral-ization. The application of FDTmay eliminate the use of sodiumhydroxide neutralizing agentsand consequently greatly reducethe quantity of contaminated sol-ids at later process steps. Duringthe freeze drying of salt/nitricacid solutions, the nitric acid andwater vapors are transported tothe condenser leaving behind aneutral salt.14 Furthermore, theacidic condensate may be re-cycled. This aspect of FDT isbeing integrated with the currentacid-recycle effort15 in theNuclear Materials Technology(NMT) division.

Application of FDT to NuclearMaterialsFDT has been applied to the de-contamination of waste streamscontaining nuclear and hazard-ous materials with great success.Acidic solutions containing ura-nium, iron chromium, andnickel originating from nuclearpower plant operations havebeen decontaminated such thatthe concentration of radioactivematerials was 1 pCi per liter(four times lower than currentEnvironmental ProtectionAgency [EPA] emission levels)while the decontamination fac-tor was 60 million.5 (The decon-tamination factor is equal to theratio of the initial and final con-centrations.) Waste solutionscontaining sodium and potas-sium nitrates, cesium-137, stron-tium-90, yttrium-91, cerium-144,ruthenium-106, and iodine-131were reduced to dry residueswhile the condensate was de-contaminated by 10 milliontimes.6 Other radioactive ele-ments, such as iodine,7 pluto-nium,8 and ruthenium,9 havebeen concentrated in the freeze-dried solid residues to produceradiochemically pure water.Organic solvents have beenseparated from radioactive sol-ids.10 Tritium oxide has beenseparated from solids, but, inthat case, the condensate con-tained the radioactive fraction.Zinc, cadmium, molybdenum,selenium, ruthenium, iron, chro-mium, and nickel also have beenseparated from radioactive sol-ids.11,12 Pilot plant-scale opera-tions for the concentrationof radioactive materials(throughput of 25 L per hour)with decontamination factorsof 100 thousand have also beenreported.13

47


A system occupying an area ofabout 225 sq ft would be capableof processing 375,000 kg peryear (based on 1500 L per dayand 250 days per year) whilemaintaining a decontaminationfactor greater than 100 million.More conventional approacheswould require considerablylarger space investments toachieve equivalent performancecharacteristics.

Plutonium FacilityDemonstrationThe FDT demonstration in thePlutonium Facility consists oftwo phases:

1) studies using a freeze drying apparatus in a glove box to demonstrate performance and to obtain design criteria for a plant-scale system, and2) integration of FDT into the existing Plutonium Facility liquid waste processing line.

In the first phase, data will beacquired using mock plutoniumsolutions for the purpose of de-termining extreme performancecharacteristics. With these ex-trema known, tests will be con-ducted using aliquots of realsolutions obtained from varioussteps along the Plutonium Facil-ity processing line. This workwill enable process engineers atthe Facility to determine the bestplace(s) to employ FDT and atwhat scale.

The integration of FDT into thePlutonium Facility waste pro-cessing line, the second phase,will require engineering andmanufacturing capabilitiesbeyond those available atLos Alamos. To meet this engi-neering and manufacturing

need, Los Alamos has enteredinto a CRADA (cooperative re-search and development agree-ment) with Edwards FreezeDrying, a commercial manufac-turer of plant-scale freeze dryers(1500 liters capacity per unit perday) based in upstate New York.Edwards Freeze Drying willwork with the NMT staff in thedesign, fabrication, and installa-tion of the glove-box system toensure that the operating param-eters can be scaled to a plant-sized system.

In this partnership, both partieshave the opportunity for sub-stantial benefit and gain.Edwards Freeze Drying consid-ers NMT Division’s knowledgeand experience with nuclearmaterials and the PlutoniumFacility to be vital for the devel-opment of this new FDT appli-cation. They see this partnershipas the foundation for the devel-opment of a new product lineand possibly new worldwidemarkets. The successful demon-stration of FDT at the PlutoniumFacility is crucial to meeting theLaboratory’s immediate andlong-range discharge goals.Furthermore, particular nuclearwaste problems facing otherDepartment of Energy (DOE)facilities such as Rocky Flats,Savannah River, and Hanfordare also being considered in thedemonstration of this technol-ogy, and solutions to those prob-lems will be integrated into theEdwards Freeze Drying productdevelopment. The DOE willgain indispensable technologyfor the environmental restora-tion of its nuclear facilities, andthe nuclear industry in generalwill have at its disposal the tech-nology to meet discharge limitsthat are becoming increasinglymore stringent.


In order to understand how FDTmight be integrated into the Plu-tonium Facility’s waste streamprocessing line, the plutoniumrecovery operation from scrapis used as an illustration.16

(See References 17–19 for morein-depth descriptions of thewaste stream processing line atthe Plutonium Facility.) Pluto-nium scrap is defined as anymaterial that has a plutoniumcomponent; usually other metalsare present. The plutoniumscrap is dissolved in either anacid or base depending on thechemical composition of thescrap. This solution enters a pro-cessing line where other metalsare separated by methods suchas ion exchange or precipitation.Three effluents are dischargedfrom this processing line. Eachvaries in the concentration of theradioactive and nonradioactiveelements. The plutonium con-centrations for the three linesare typically 0.05, 0.08, and 0.15 gplutonium per liter. Usually theradioactive component is a smallfraction of the total volume, buteven these small amounts are notallowed to be discharged into theenvironment.

In an effort to further concentratethe radioactive component andto pass an effluent that has alower radioactive concentration,the three effluents exiting fromthe ion exchange and precipita-tion process are sent to an evapo-rator. The evaporator effluenthas a concentration of about7 x 10-7 g plutonium per literand is sent to the Liquid WasteTreatment Facility (TA-50) at LosAlamos for further processing.


Fig. 2. Configurations for plutonium effluent processing line: (a) currentprocessing line for plutonium scrap at the Plutonium Facility and perfor-mance characteristics, (b) proposed simple substitution of a freeze dryerand its performance characteristics, (c) proposed evaporator and freezedryer in series and their performance characteristics.



The residues after evaporation(called bottoms) are stabilized incement. The cements are sentoff-site for disposal and/orstorage. Figure 2A illustrates theevaporator feeds, evaporatorproducts, and storage anddisposal paths.

Two of the possible applicationpoints for FDT in the processingline are (1) as a replacement forthe evaporator or (2) in seriesafter it. Either of these applica-tions may also feature a meansfor stabilizing the solid residuesthat uses atomically mixedceramic or cement precursors(see next section). Figures 2Band 2C illustrate these FDTapplication configurations. Ananalysis and comparison sum-mary of evaporator performanceand the performance of configu-rations using FDT is given inTable I and discussed in thefollowing paragraph. Of thethree evaporator feeds, thesolutions with a concentrationof 0.15 g plutonium per literrepresent only a small fraction

of the total waste pro-cessed; the largestfractions contain only0.05 – 0.08 g plutoniumper liter. Because therelative fractions arevariable over time, theworst case scenariowas chosen for the

performance analysis; that is,we assumed that 100% of thefeed has a concentration of 0.15 gplutonium per liter.

49

Table I. Comparison of Present Evaporator Performance to FDT and to Their Performance in Series.a



have to be sent to TA-50, andthis could have a significantpositive impact on that wastefacility’s operations as well. Incalendar year 1991, the pluto-nium-239 effluents from TA-50averaged 43 pCi per liter.20

Current decontamination factorsat TA-50 are on the order ofabout 1000 (based on the aver-age 1991 monthly influents andeffluents for all radioactivematerials). Clearly, if FDTenjoyed a successful demonstra-tion at the Plutonium Facility,this success would imply thatdecontamination factors atTA-50, as well as at other DOEfacilities, would be lower (byfactors in excess of 1000) thancurrent EPA discharge limits.In conjunction with the prepara-tion of atomically mixed ceramicprecursors (see the next section),FDT may prove dually advanta-geous: first, the liquid wasteswould be greatly decontami-nated; second, the solid residuesmay be more easily stabilized,

which would result in the simpli-fication or elimination of somepost-evaporator operations.

Stabilization of RadioactiveSolids Using Freeze-DriedAtomically Mixed PrecursorsA freeze drying process for thepreparation of ceramic precur-sors that are atomically mixedhas been developed and provento be very useful for overcomingkinetic and thermodynamicbarriers encountered duringconventional ceramic materialsprocessing. By integrating theseparation aspects of FDT withthe preparation of atomicallymixed precursors (AMPs), theradioactive solid residues fromthe separation process can betransformed into materials withexceptional environmental stabil-ity. These materials can besynthesized faster, at lowertemperatures and pressures, andin a fashion compatible with theglove-box environment on anindustrial scale.

Using FDT as a replacement forthe evaporator at the PlutoniumFacility and assuming a decon-tamination factor of 100 million,the liquid effluents wouldcontain 1.5 x 10-9 g plutoniumper liter or 93 pCi per liter.A decontamination factor of 100million is not at all optimisticbecause other developmentsmade in conjunction with ourindustrial partner have indi-cated that decontaminationfactors far in excess of onebillion should be achievable.(The details of these develop-ments were proprietary at thetime of the writing of this paperand could not be disclosed.)If a freeze dryer is placed inseries and downstream from the evaporator at the PlutoniumFacility, the effluent wouldcontain 7 x 10-12 g plutoniumper liter, which is 0.4 pCi perliter or 10 times lower than thecurrent EPA discharge limits.In that case, discharges may not

Present Evaporator FDT Performanceb Evaporator and FDTb

Performance (alone) (in series)

(g Pu l-1) (dpm l-1)c

(g Pu l-1) (dpm l-1) (g Pu l-1) (dpm l-1)

Input 0.15d

2.0 x 1010 0.15d

2.0 x 1010 0.15d

2.0 x 1010

Output 7.0 x 10-4d9.3 x 107 1.5 x 10-9 200 7.0 x 10-12 1.0

Decontamination 210 210 108 108 2 x 1010 2 x 1010

Factor

aEvaporator is placed upstream relative to a freeze dryer.bIn light of new proprietary developments, the actual decontamination factors may be substantially greater (1000 times)

than those cited here.c Disintegrations per minute per liter, assuming a plutonium-239 half-life of 24,100 years.d Maximum input values and minimum output values were chosen from Ref.16; in practice, actual values

will probably be lower by a factor of 3 or greater.


Fig. 3. Schematic for the preparation of atomically mixed precursors ofeither ceramics or cement, which contain the solid, freeze-dried residues.Using this approach, solid residues would be rendered more stable as wellas liquid wastes decontaminated.



compared to that of either boro-silicate glasses or supercalcinematerials,24 but the high tem-peratures and pressures re-quired for their synthesis havemade these materials unattrac-tive as a waste form. Becauseof the unique characteristics ofAMPs, we are applying them tothe stabilization of uranium-and plutonium-containing solidsresulting from the FDT decon-tamination process. There aremany crystalline phases intowhich uranium, plutonium, andother radioactive elements canbe incorporated.25 High tem-peratures and/or pressures maybe eliminated. Cements with adispersion of the radioactivecomponent that is homogeneouson nanometer-length scales,without the use of mechanicalmixers, may also be prepared.Formulations based on existingmineral or cement compositionsusing chemical surrogates foractinide materials, such as thosefound at the Plutonium Facility,are currently being developedand the products characterized.

Atomic mixtures are a mixtureof two or more componentslacking both long- and short-range structural and composi-tional order. Atomic mixturesare prepared by rapidly freezing(dT/dt = one million °C per sec-ond) a solution containing theappropriate salts followed byfreeze drying.14 (See Fig. 3 for adiagram of this freeze dryingprocess.) The high freezing ratestrap solute molecules in a disor-dered state in a matrix of ice.Then, by freeze drying, the icematrix is removed, and the dis-ordered state of the original so-lution is preserved in theresidual solid. These atomicmixtures are extremely reactive,and the kinetics for product for-mation are unlike those of con-ventional ceramic processing.A microscopic picture for the be-havior of these mixtures is beingdeveloped in conjunction withthe staff from several otherLaboratory divisions* usingtime- and temperature-resolvedneutron and x-ray scattering andMonte-Carlo simulations of themolecular dynamics. The cal-cine temperature of AMP mate-rials has been shown to be lower(by greater than 100 °C)21 thanthose of conventional ceramicprocesses, while the reactiontimes are also drasticallyreduced.

As indicated in Fig. 2A, cementsare currently mixed with theradioactive evaporator bottomsfor the purpose of stabilization.While this composite allows forsafe temporary storage andtransportation, these cements

may not be considered suitablefor long-term storage because ofnew performance requirementsimposed by regulating agencies(EPA) and are not considered tobe stable on time scales of geo-logical porportions. Other mate-rials such as glasses and mineralanalogues, for example synroc,are, or have been, candidates forlong-term, permanent storage ofnuclear materials, but these can-didates also have problems.Although the manufacturingprocess for glasses is compatiblewith the radiological environ-ments and procedures, seriousdisadvantages exist; for ex-ample, borosilicate glass ismetastable with respect to acrystalline-phase assembly ofthe same composition22 and,thus, readily devitrifies whensubjected to elevated tempera-tures and pressures.23 Crystal-line materials such as synrochave been shown to have far su-perior environmental stability

* Participants include Los AlamosNeutron Scattering Center(LANSCE) and the MaterialsDivision.

51


17. D. C. Christensen et al.,“Waste from Plutonium Conversionand Scrap Recovery Operations,”Los Alamos National Laboratoryreport, LA-11069-MS (March 1988).

18. S. L. Yarbro, “Using SolventExtraction to Process Nitrate AnionExchange Column Effluents,” LosAlamos National Laboratory report,LA-11007-T (October 1987).

19. E. L. Christensen and W. J.Maraman, “Plutonium Processingat the Los Alamos Scientific Labora-tory,” Los Alamos Scientific Labora-tory report, LA-3542 (April 1969).

20. R. T. Harris, “Summary ofOperating Data,” in Annual Report,EM-7 Liquid Waste ManagementGroup, 1991 (approved October1992), Los Alamos National Labora-tory document.

21. N. V. Coppa, et al., J. Mater. Res.7, 2017 (1992).

22. G. J. McCarthy, Nuclear Technol-ogy 32, 92 (1977).

23. G. J. McCarthy et al., Nature273, 216 (1978).

24. A. E. Ringwood et al., Nature278, 219 (1979).

25. G. J. McCarthy, “Ceramics andGlass Ceramics as High LevelWaste Forms,” in Ceramic & GlassRadioactive Waste Forms, D. W.Ready and C. R. Cooley, Eds. (Dept.of Energy and Research Develop-ment, Washington, D.C., 1977),Conf. 770102, pp. 83-99.


The synthesis and characteriza-tion of actinide-containing mate-rials are planned for the nearfuture. Research using AMPsthat contain nuclear materialsis focused on the developmentof several high-performanceceramics and cements, but thisresearch will also advance thefrontier of multication, solid-state actinide materials.♦

Cited References1. J. D. Mellor, Fundamentals ofFreeze Drying (Academic Press,London, 1978).

2. R. B. Bird, W. E. Stewart, andE. N. Lightfoot, Transport Phenomena(John Wiley & Sons, New York,1960), pp. 563 - 572.

3. T. R. Sheng and R. E. Peck,“Rates for Freeze Drying,” in WaterRemoval Processes: Drying andConcentration of Foods and OtherMaterials, C.J. King and S. P. Clark,Eds., AICHE Symposium Series 73(163), pp. 124-130 (1977).

4. E. W. Flosdorf, Freeze Drying(Reinhold, New York, 1949),pp. 1-64.

5. O. Takanobu and I. Kondoh,“Method and Apparatus forTreatment of Liquid RadioactiveWaste,” Ger. Offen., European PatentNo. 3,625,602 (1987).

6. N. V. Mikshavich et al., At. Energ.27(6), 541 (1969).

7. K. Ohtsuka, J. Ohuchi, and T.Suzuki, “Method of RecoveringRadioactive Iodine in Spent NuclearFuel Retreatment Process,” Euro-pean Patent Application No.361,773 (1990).

8. K. Ohtsuka, I. Kondoh, andT. Suzuki, “Spent Fuel TreatmentMethod,” European Patent Appli-cation No. 358,431 (1990).

9. S. Tachimori, J. Nucl. Sci. Technol.13, 442 (1976).

10. G. E. Harwood et al. ,”Process-ing of Liquid Scintillation Wasteand Animals,” in Synthesis andApplication of Isotopically LabeledCompounds, Proceedings of Interna-tional Symposium, W. P. Duncanand Alexander B. Susan, Eds.(Elsevier, Amsterdam, 1983),pp. 117-120.

11. Y. Wang, J. Qin, Q. Ji, S. Wu,and X. Wang, Fenxi Huaxue 13, 210(1985).

12. K. Ohtsuka et al., “FreezeDrying in Reprocessing NuclearReactor Fuels,” Jpn. Kokai TokkyoKoHo, Japan Patent No. 1,316,695(1989).

13. K. Jessnitzer et al., Nucl. Sci.Abstr. 23(2), 2542 (1969).

14. N. V. Coppa,”Synthesis andCharacterization of High TransitionTemperature Superconductors,”Ph.D. thesis, Temple University(1990), pp. 71-134.

15. T. R. Mills, “Acid Recovery andRecycle,” this report.

16. S. L. Yarbro, Los AlamosNational Laboratory, privatecommunication, September 1993.



Nuclear Applications for Magnetic SeparationsLarry R. Avens,Laura A. Worl,Ann R. Schake,Dennis D. Padilla,and Karen J. deAgueroNuclear Materials Processing:Chloride Systems

Kevin B. Ramsey,Richard D. Maestas,Robert C. Martinez,and Yvette E. ValdezNuclear Materials Processing:Nitrate Systems

Thomas L. ToltLockheed Environmental Systems andTechnologies Company

F. Coyne Prenger,Walter F. Stewart,and Dallas D. HillAdvanced Engineering TechnologyEngineering Sciences andApplication Division

Table I. Volume Magnetic Susceptibility ofSelected Compounds and Elements

*SI Units

Magnetic Separation ProcessMagnetic separation is a physical separation process that segregatesmaterials on the basis of magnetic susceptibility.1 Because the pro-cess relies on physical properties, separations can be achieved whileproducing a minimum of secondary waste.

When a paramag-netic particle en-counters a non-uniform magneticfield, the particle isurged in the direc-tion in which thefield gradientincreases. Diamag-netic particles reactin the oppositesense. When thefield gradient isof sufficiently highintensity, paramag-netic particles canbe physically cap-tured and separatedfrom extraneousnonmagneticmaterial.

Because all actinidecompounds areparamagnetic(Table I), magneticseparation of ac-tinide-containingmixtures is feasible.Magnetic separationon recycle pluto-nium chemical pro-cess residues hasbeen demonstratedon an open-gradientmagnetic separator.2The advent of reli-able superconduct-ing magnets alsomakes magneticseparation of weak-ly paramagneticspecies attractive.

Nuclear Materials Technology 1993 Report

“Magnetic separation is aphysical separation processthat segregates materials onthe basis of magneticsusceptibility.”

Compound/Element Susceptibility x 106*

FeO ........................... 7178.0Fe

2O

3. ....................... 1479.0

UO2. .......................... 1204.0

Cr2O

3. ........................ ..844.0

NiO ............................ ..740.0

Am ............................. ..707.0

Pu .............................. ..636.0

U................................ ..411.0

PuO2. ........................ ..384.0

CuO........................... ..242.0

RuO2. ........................ ..107.0

Th .............................. ....41.0

UO3. .......................... ....41.0

CrO3. ......................... ....14.0

Si ............................... .....-0.3

CaO........................... .....-1.0

ZrO2. ......................... .....-7.8

MgO .......................... ...-11.0

KCl ............................ ...-13.0

CaCl2. ........................ ...-13.0

NaCl .......................... ...-14.0

CaF2. ......................... ...-14.0

SiO2. .......................... ...-14.0

MgF2. ........................ ...-14.0

Graphite .................... ...-14.0

Al2O

3. ........................ ...-18.0

53

Magnetic Separation MethodsAlthough numerous magneticseparation methods exist, wehave currently selected twomethods for development: mag-netic roll (or drum-type) separa-tion and high-gradient magneticseparation (HGMS). The useful-ness of the open-gradient sepa-ration method used in the earlywork is limited by its low pro-cessing rate.2

The magnetic roll separationmethod is used to separatemagnetic particles from drypowders. A diagram of a rollseparator is shown in Fig. 1.

In practice, a dry powder is de-livered onto a thin belt thatmoves over rollers. The frontroller is fabricated from a neo-dymium-iron-boron rare earthmagnet. Ferromagnetic and suf-ficiently paramagnetic particlesare attracted to the magnet andadhere to the belt in the regionof the magnet. As the belt movesaway from the magnet, the fer-romagnetic and paramagneticparticles disengage from the beltand are collected in a catch pan.Diamagnetic and nonmagneticparticles pass over the magneticroller relatively unaffected andare collected in a different catchpan. This method effectively ex-tracts particle sizes ranging from90 to 850 µm.

Fig. 1. Diagram of a roll magnetic separator.

The HGMS method is used toseparate magnetic componentsfrom solids, liquids, or gases. Adiagram of the method is shownin Fig. 2. Most commonly, thesolid is slurried with water andpassed through a magnetizedvolume. Field gradients are pro-duced in the magnetized vol-ume by a ferromagnetic matrixmaterial, such as steel wool, ironshot, or nickel foam. Ferromag-netic and paramagnetic particlesare extracted from the slurry bythe ferromagnetic matrix, whilethe diamagnetic fraction passesthrough the magnetized volume.The magnetic fraction is flushedfrom the matrix later when themagnetic field is reduced to zeroor the matrix is removed fromthe magnetized volume. TheHGMS method effectively ex-tracts particle sizes rangingfrom ~0.3 to 90 µm and, thus,is complementary to the rollseparator.


Nuclear Applications for Magnetic Separation


A second 3-inch-bore supercon-ducting magnet, installed at ourwaste treatment facility, will beused to study waste streamtreatment applications andwill eventually be interfacedto a glove box in the PlutoniumFacility for tests of chemicalprocessing residue concentra-tion. Finally, a 6-inch-boremagnet, installed in an openarea, is being used for nonra-dioactive tests and for prototypedevelopment of soil decontami-nation applications.

High-Gradient MagneticSeparation (HGMS) ModelMore than 10 variables indepen-dently affect the HGMS process;thus, an analytical model for theprocess is needed to select ap-propriate bench-scale experi-ments and to effectively analyzethe resulting data. In addition, avalidated analytical model sup-ports prototype design and pro-cess scale-up. Much work6-12

by others has been done tomodel the behavior of the para-magnetic particles as they inter-act with the magnetized matrix.Areas investigated include thedynamic effects of particle tra-jectories in homogeneous mate-rials, the influence of particlebuildup on the matrix elements,and the particle buildup effecton the flow field. These studies,though requiring simplifyingassumptions, identify appropri-ate independent variablesthat form the basis for thisinvestigation.

Fig. 2. Simplified HGMS diagram.

The development of the HGMSmethod has been the centralfocus of Los Alamos NationalLaboratory’s magnetic separa-tion research.3-5 Today’s super-conducting magnet technologyproduces much higher electro-magnetic fields than is possiblewith conventional electro-magnets, which are limited bythe saturation of iron to a mag-netic field strength of about twotesla. This availability of highfields permits a broad range ofHGMS applications, including

liquid waste stream treatment,underground storage-tank wastetreatment, chemical processingresidue concentration, and soildecontamination.

Our current experimentalHGMS equipment includes oneconventional-coil separator andthree superconducting high-gra-dient separators. We are usingthe 1-inch-bore conventional coiland one 3-inch-bore supercon-ducting magnet, installed on avent hood in the Plutonium Fa-cility, for low-activity samplessuch as contaminated soils anduranium-containing surrogates.



55

= Capture heightElement diameter

Capture heightMatrix element

Flow

Dm Dm

Viscous drag

Magnetic

Gravity

Matrixelement

Paramagneticparticle

Flowstream

We propose a rate model for thisprocess that depends on a sepa-ration coefficient, which we de-fine in terms of a capture crosssection and a potential functiondefined by the force balance onthe particle. A series of tests

Fig. 3. Force balance of magnetic, viscous drag, and gravity forces on aparamagnetic particle in the vicinity of a matrix element.

Fig. 4. The matrix capture cross section defined as the ratio of captureheight to matrix element diamter Dm.

ParamagneticParticle

The matrix materials currentlyused for HGMS are inhomogen-eous and have a complex crosssection. In addition, the para-magnetic particles arenonspherical and include arange of particle sizes. All ofthese factors preclude preciseanalytical treatment. If the par-ticles are physically liberatedfrom the host material and arenot electrically charged, theprincipal forces governing theirbehavior are magnetic, viscous,and gravitational.

The performance of the mag-netic separator is modeled usinga static force balance on an indi-vidual paramagnetic particle inthe immediate vicinity of a ma-trix element, as shown in Fig. 3.The model assumes that if themagnetic capture forces aregreater than the competing vis-cous drag and gravity forces,the particle is captured andremoved from the flow stream.

ViscousDrag

MatrixElement

FlowStream

was performed to evaluate sepa-rator performance as a functionof the independent variables inthe separation process. Theresults were used to determinethe separation coefficient in theproposed rate model and werecorrelated using physicallymeaningful, dimensionlessgroups.

We assume that there exists foreach matrix element a captureheight, wherein each magneticparticle entering the window de-fined by the capture height andthe matrix element diameter willbe captured by the matrix ele-ment. We then define a capturecross section for a matrix ele-ment to be the ratio of the cap-ture height to the matrix elementdiameter, as shown in Fig. 4.The separation coefficient is re-lated to the capture cross sec-tion, and if the capture crosssection is known, the separationefficiency can be calculated.

Matrix ElementCapture Height




Table II. Independent Variables Investigated

Material Characteristics Separator Parameters

Particle Size: 0.5−50 µm Matrix Element Size: 5−100 µm

Impurity Concentration: 0.4−400 ppm Matrix Element Spacing: 80−1200 µm

Solids Concentration: 5−30 wt % Magnetic Field Strength: 0.5−7.5 T

Volume Magnetic Matrix Material: 43D Stainless Steel & Ni

Susceptibility* (x106

): 129−1478

Surfactant Concentration: 0.05−0.2 wt % Residence Time: 1.0−8.0 s

Slurry pH: 7−10 Superficial Velocity: 0.25−4.0 cm/s

*SI units

HGMS Data CorrelationA comprehensive series of ex-periments was performed to de-termine the capture cross sectionas a function of the independentvariables. Table II lists the vari-ables investigated and theirrange. For convenience, the listhas been divided into materialcharacteristics and separator pa-rameters. A total of 120 experi-ments were conducted; 76involved water/contaminantmixtures and 44 involved wa-ter/soil/contaminant mixtures.

The capture cross section foreach test was determined experi-mentally based on the measuredfeed and effluent concentrations,the particle force balance, thematrix geometry, and the con-taminant particle size.

Intuitively, one expects to haveto maximize both the separatorlength and the matrix residencetime for optimum performance;however, only their ratio has tobe considered, and this factmakes superficial velocity themore fundamental parameter.

HGMS ResultsWe have completed a compre-hensive series of HGMS experi-ments with nonradioactivesurrogates, as discussed in theprevious section, and have pro-gressed to tests with radioactivematerial. The results to date arevery promising.

The capture cross section is sig-nificantly less than 1.0, whichimplies that particles are sweptaround the matrix elements bythe fluid flow even if the par-ticles enter the region defined bythe projected frontal area of thematrix element. In addition, thecapture cross section is propor-tional to the magnetic force andthe contaminant particle diam-eter and inversely proportionalto the matrix element diameterand the matrix element spacing.These results are consistent withour observations. The capturecross section is inversely pro-portional to the superficial ve-locity, which is the ratio of theseparator length and the slurryresidence time in the matrix.



57

To investigate chemical process-ing applications, we experi-mented with contaminatedgraphite residues. These seriesof tests revealed that HGMS per-formance is strongly influencedby the graphite solids content.Copper oxide serves as the sur-rogate for plutonium oxide.Figure 6 illustrates the effect ofmagnetic field and solids frac-tion on the separation effective-ness for particles less than 90µm. The highest separation effi-ciency is observed under condi-tions of low solids fraction(<5% solids), with a fairlyminor effect from external fieldstrength. As discussed earlier,the effective particle size rangefor the roll separator on drypowders is greater than 90 µm;thus, the coupling of HGMSwith the roll separator wouldbe advantageous for treatmentof chemical processing residues.

Our application of HGMS to theTank Waste Remediation System(TWRS) indicates that we maybe able to reduce certain types oftank waste directly to non-tran-suranic (non-TRU) levels in onestep. The initial tank waste tar-get is Neutralized Cladding Re-moval Waste (NCRW), which isformed from the chemicaldecladding of Zircaloy-cladmetallic uranium fuel that hasbeen made alkaline for storage.It contains zirconium hydrous-oxide solids, TRU elements(~1000 nCi/g in sludge), andmixed fission products. Oursurrogate components forNCRW tank waste weregenerated from PUREX reagentsused in NCRW production.13

With silica, however, ~86% ofthe plutonium in a spikedsample was removed by HGMSat two tesla. Surface adsorptionappears to cause silica sand tobind plutonium oxide particlesto the surface. We also are plan-ning HGMS soil tests with the3-inch superconducting magnetlocated in the Plutonium Facil-ity. Model predictions and sur-rogate work suggest increaseddecontamination levels at fieldstrengths above two tesla.

Passes Through Separator

1

10

100

1000

10000

Plut

oniu

m C

once

ntra

tion

(pCi

/gm

)

99.8%

Illite-Beidellite : <45 mSiO2 : >10 m <45 m

85.6%

% Extraction

0 21 31

101

102

103

104

Fig. 5. HGMS extraction of plutonium oxide from soils (external magneticfield strength of 2.0 tesla ; 2–5 µm PuO2 particles).

To investigate soil decontamina-tion applications, we workedclosely with Lockhead Environ-mental Systems and Technolo-gies in a CRADA (cooperativeresearch and developmentagreement) effort. Our first testswith a spiked clay-type soilshowed almost three orders ofmagnitude decontamination —from about 1500 pCi/g to lessthan 4 pCi/g with an appliedfield of two tesla (see Fig. 5).




Prior to precipitation of thehydrous oxides, uranyl nitrateor iron nitrate was added to theacidic solution serving as theNCRW TRU surrogate. HGMStests of aged and unaged slur-ries containing precipitated ironand zirconium have been con-ducted at a one-tesla field, andresults indicated that 84% ofthe aged ferric hydrous oxidescan be captured from a slurry.Because of this favorable out-come using the iron solutions,HGMS tests are planned on zir-conium/uranium/hydrous-oxide solutions.

Magnetic Roll SeparationResultsA variety of hot tests have beenconducted on a lab-scale rollseparator in a glove box. Sev-eral lots of graphite powder,bomb reduction sand (MgO),and sand, slag, and crucible(SS&C) residues have been pro-cessed with variable results.The best results were obtainedwith graphite processing, inwhich 85% of the plutoniumwas concentrated into 4% of thebulk material. Particles weresized greater than 125 µm. Theplutonium content of the leanfraction was sufficiently low inplutonium so that it could bediscarded directly to grout.

The roll separator results werenot as favorable with MgO orSS&C residues. In the best casefor MgO, 68% of the plutoniumwas concentrated in 44% of thebulk material. For SS&C, the re-sults were inconclusive with nodiscrete separation. Yet, in thepast, we have shown that mag-netic separation can achievehigher performance with theseresidues; for example, with theopen-gradient magnetic separa-tor, we were able to concentrate81% of the plutonium in 33% ofthe SS&C bulk material.2

We feel that, if we can increasethe magnetic field intensity, aviable roll separation technologycan be developed. The magneticfield intensity acting on the feedparticles in magnetic roll separa-tion is inversely proportionalto the thickness of the belt.

HGMS is also being developedfor magnetic filtration of fluidwaste streams generated duringactinide chemical processingand for those generated at thewaste treatment facility. Theanalytical model predicts thatHGMS can reduce the actinideconcentration in liquid wastestreams by several orders ofmagnitude.

Solids Fraction Magne

tic F

ield,

B(T

)

Effe

ctiv

enes

s

2

0.5

1

1.5

0.1

0.2

0.3

0

0.5

1

Fig. 6. HGMS effectiveness for contaminated graphite slurries as afunction of magnetic field strength and graphite solids fraction.The plutonium surrogate was 4- m copper oxide particles, andthe graphite solids were sized < 90 m for the experiments.



59

If the separation efficiency ishigh, the secondary componentcan be discarded, and the pri-mary component, with reducedvolume, can be further pro-cessed or packaged for disposal.Concentration of the actinidesfrom extraneous materials be-fore further processing yieldsmore efficient recovery or treat-ment operations; for example,concentrating the contaminantsreduces the volume of chemicalreagents (acids) necessary forsubsequent operations. Further-more, increased concentration ofcontaminants in dissolved feedsallows a more efficient ion ex-change or solvent extraction unitoperation. Because less extrane-ous material is leached anddissolved, the salt load on sub-sequent waste treatment opera-tions is reduced.

Magnetic separation could beapplied to on-site decontamina-tion of radioactive spills. How-ever, development of a mobilemagnetic separation unit wouldbe required. Portable supercon-ducting magnet systems do existfor mobile Magnetic ResonanceImaging. Development of mag-netic filtration may be a methodto cut to very low levels the TRUeffluents from Department ofEnergy facilities and, thereby,greatly reduce waste treatmentcost. ✦

In the glove-box environment,belt changes and roller align-ment are very difficult with ourcurrent roll separator. A secondroll separator of improved de-sign that uses a very thin (0.005inch) stainless-steel belt hasbeen obtained. This separatorwill be tested in a nonradioac-tive environment and modifiedfor convenient glove-box use.

Benefits of MagneticSeparationWe have shown that HGMS,open-gradient magnetic separa-tion, and magnetic roll separa-tion can be used to concentrateplutonium and uranium inwaste residues and contami-nated soils. One of the majorbenefits of magnetic separationtechnology is that it only parti-tions the existing waste volume.A high contaminant concentra-tion of a primary componentand a low contaminant concen-tration of a secondary compo-nent are created.



Cited References1. W. J. Lyman, “High-GradientMagnetic Separation,” in Unit Op-erations for Treatment of HazardousIndustrial Wastes, D. J. De Renzo,Ed. (Noyes Data Corporation, ParkRidge, New Jersey, 1978), pp. 590–609.

2. L. R. Avens, U. F. Gallegos, andJ. T. McFarlan, “Magnetic Separa-tion as a Plutonium Residue En-richment Process, Separation Scienceand Technology 25, 1967 (1990).

3. L. R. Avens et. al, “MagneticSeparation for Soil Decontamina-tion,” Waste Management Confer-ence ’93, Tucson, Arizona, February28−March 4, 1993, Los AlamosNational Laboratory report LA-UR-93-229.

4. L. A. Worl et. al, “Remediationof Hanford Tank Waste Using Mag-netic Separation,” Waste Manage-ment Conference ’93, Tucson,Arizona, February 28−March 4,1993, Los Alamos National Labora-tory report LA-UR-92-3977.

5. F. C. Prenger et. al, “High Gradi-ent Magnetic Separation Applied toEnvironmental Remediation,”Cryogenic Engineering Conference,Albuquerque, New Mexico, July12–16, 1993, Los Alamos NationalLaboratory report LA-UR-93-2516.


11. Z. J.J. Stekly and J. V. Minervini,“Shape Effect of the Matrix on theCapture Cross Section of Particlesin High Gradient Magnetic Separa-tion,” IEEE Transactions on Magnet-ics Mag-12, 474–479 (1976).

12. J. H. P. Watson, “Magnetic Fil-tration,” Journal of Applied Physics44, 4209–4213 (1973).

13. “PUREX Technical Manual,PUREX Systems and TechnologyChemical Processing System Engi-neering,” Westinghouse HanfordCompany report WHC-CM-5-25(July 1988).

6. I. V. Akoto, “Mathematical Mod-eling of HGMS Devices,” IEEETransactions on Magnetics Mag-13,1486–1489 (1977).

7. F. J. Friedlander, M. Takayasu, J.B. Retttig, and C. P. Kentzer, “Par-ticle Flow and Collection Process inSingle Wire HGMS Studies,” IEEETransactions on Magnetics, Mag-14,1158–1164, (1978).

8. W. F. Lawson, W. H. Simons, andR. P. Treat, “The Dynamics of a Par-ticle Attracted by a MagnetizedWire,” Journal of Applied Physics 48,3213–3224 (1977).

9. F. E. Luborsky and B. J.Drummond, “Buildup of Particleson Fibers in a High-Field, High-Gradient Separator,” IEEE Transac-tions on Magnetics Mag-12, 463–465(1976).

10. J. E. Nesset and J. A. Finch, “AStatic Model of High-GradientMagnetic Separation Based onForces within the Fluid BoundaryLayer,” in Industrial Applications ofMagnetic Separation, Y. A. Liu, Ed.(Institute of Electrical and Elec-tronic Engineers, New York, 1979)pp. 188−196.



61


Distillation Separation of Actinides from Waste Salts

Eduardo Garcia,Vonda R. Dole,Walter J. Griego,John J. Lovato,James A. McNeese,and Keith AxlerNuclear Material Processing:Chloride Systems

IntroductionPlutonium-bearing residue salts resulting from past pyrochemicalprocessing have accumulated in the Department of Energy (DOE)weapons complex. This problem is especially acute at Rocky Flats,where approximately 20 metric tons of salt await disposition.Pyrochemical salt wastes will also be generated in the future byComplex 21. In addition to containing plutonium with its attendanthazards, these salts may contain reactive metals. The reactive com-ponent places the salts in the category of mixed waste and unac-ceptable under Waste Isolation Pilot Plant (WIPP) acceptancecriteria. Because of restrictions on plutonium loading in individualwaste drums, WIPP may not even have the capacity for the numberof drums that would be generated, even if the waste were accept-able. The salts must therefore be treated to remove as much pluto-nium as possible and render inert any reactive components.

Physical separation processes, such as distillation separation, aremore attractive than chemical or dissolution processes becausephysical processes generate much less secondary waste. A physicalseparation of waste salts from plutonium is theoretically possiblebecause of a large difference in vapor pressure. A high-temperaturedistillation process can be used to volatilize the alkali and alkaline-earth chloride salts from essentially nonvolatile actinide oxides.An oxygen sparging process developed at Los Alamos pretreatsthe salts to convert actinide species to oxides or oxychlorides andsimultaneously converts reactive metals into inert oxides so thatthey are no longer mixed waste. The salt is then evaporated toleave a plutonium dioxide heel suitable for storage. Waste volumewill be essentially limited to the salts themselves. Moreover, thegreatly reduced plutonium content of the salts will result in agreatly reduced number of waste drums for WIPP disposal and alarge reduction in the cost of disposal. It is theoretically possibleto remove plutonium below the 1 ppm level, which would thenclassify the salts as low-level waste (LLW) rather than transuranic(TRU) waste and result in even greater cost savings and a morebenign waste form.

“It is theoretically possibleto remove plutonium belowthe 1 ppm level, whichwould then classify the saltsas low-level waste ratherthen transuranic waste andresult in even greater costsavings and a more benignwaste form.”


Distillation Separation ProcessThe distillation separation process can be modeled by the kinetictheory of gases.1 From this theory, one can show that the rate of themass of a gas incident on a unit area per unit time is

, (1)

whereW = gas incident rate (g cm-2 sec-1),M = molecular weight (g mol-1),R = gas constant (8.325 x 107 ergs K-1 mol-1),T = temperature (K),P = vapor pressure (torr) at temperature T, andC = conversion factor (1333 dyne cm-2 torr-1).

Substituting the values for the gas constant and the conversionfactor produces the rate equation

. (2)

If the temperature of the surfacethat the gas strikes is muchcolder than the temperature ofthe gas vapor, then condensationwill occur with a sticking coeffi-cient of 1. Under these condi-tions, the condensation rate isequal to the incident rate, and,thus, the amount of materialdeposited can be estimated byusing the calculated incidentrate. In the case of molten saltdistillation, the condenser tem-perature will be much lowerthan the temperature at whichthe material is evaporating, sothat estimating the amount de-posited will be possible. Mostimportantly, calculations can bemade for the rate of depositionof pyrochemical salts becausethe vapor pressure of the vari-ous components, with one ex-ception, is known.

Table I lists vapor pressures ofcommon pyrochemical salt com-ponents at selected tempera-tures.2-9 The extremely lowvapor pressures for plutoniumdioxide and calcium fluoride areextrapolations from higher tem-perature data. Table I showsthat the best separation can beachieved between plutoniumdioxide and the chloride salts;it also shows that a good separa-tion between the salts and pluto-nium trichloride cannot beachieved. To ensure optimalseparation, Los Alamos pretreatsthe pyrochemical waste salts us-ing an oxygen sparging process,or alternate oxidation processes,to convert plutonium and ameri-cium species to oxides oroxychlorides.

Compound 850 °C 950 °C 1050 °C Ref. No.

NaCl 10-0.063 10.0.64 101.2 2

KCI 100.23 100.90 101.5 3

MgCl2 100.27 100.97 101.6 4

CaCl2 10-2.9 10-1.9 10-1.0 4

CaF2 10-7.7 10-6.1 10-4.7 5

Pu 10-8 10-7 10-5 6

PuCl3 10-1.8 10-0.86 10-0.086 7

PuO2 10-15.7 10-13.5 10-11.7 8

PuOCl 10-6 10-5 9

Table I. Vapor Pressure of Pyrochemical Salt Components



63

Table II lists the rate of deposi-tion (at selected temperatures)of various compounds. Figure 1shows calculated depositionrates as a function of tempera-ture. These are calculated rates,however, and actual rates canbe slower by some orders ofmagnitude. Nevertheless, all

the chloride salts, except cal-cium chloride, can be distilledat acceptable rates below900 °C. Calcium chloride canbe distilled at 1100 °C. Even at1100 °C, the vapor pressure ofplutonium and americiumdioxide is low enough to stillget an excellent separationfrom the salt.

Fig. 1. Distillation rates as a function of temperature.

Table II. Deposition Rate (g hr-1) from 100 cm2 Surface Area

Compound 850 °C 950 °C 1050 °C

NaCl

KCI

MgCl2CaCl2CaF2

AmO2

PuO2

Pu

The deposition rates can be usedto calculate the composition ofthe distillate salt under idealconditions. For example, in asalt that contains only sodiumchloride and plutonium dioxideat 850 °C with a 100 cm2 surfacearea, the rate of deposition of so-dium chloride is 4100 g per hourwhile that of plutonium dioxideis 10-12 g per hour. The distillatesalt would then contain a weightconcentration of 10-10 ppmplutonium. At 1050 °C, the dis-tillate salt would have a pluto-nium concentration of 10-7 ppm.The LLW criterion for pluto-nium-239 is on the order of1 ppm. Therefore, vacuum dis-tillation of a sodium chloride,potassium chloride, or magne-sium chloride salt that containsonly plutonium dioxide has thepotential to produce a waste saltwith a plutonium concentration10 orders of magnitude belowthe LLW criterion. In the case ofcalcium chloride, the calculatedconcentration is 10-5 ppm at1050 °C. Figure 2 shows a plotof plutonium concentration invarious distilled pyrochemicalsalt components as a function oftemperature. As shown in thisplot, even at 1200 °C all the chlo-ride salts will be considerablybelow the 1.6 ppm LLW cutofffor plutonium. The americiumcontent must be below 0.03 ppmfor LLW, and this requirementcan also be met assuming all theamericium is present as AmO2.

4100 20000 75000

9100 41000 140000

11000 55000 200000

9 85 560

10-3.9 10-2.3 10-1.0

10-11 10-9.0 10-7.2

10-12 10-9.6 10-7.8

10-3.7 10-2.4 10-1.4




Fig. 2. Calculated plutonium concentration in distilled salts assuming only PuO2 present.

A picture of the waste salt beforevacuum distillation is shown inFig. 3, and the resulting distilledsalt and plutonium-dioxide heelare shown in Fig. 4.

Preliminary experiments withalternate oxidation pretreat-ments have also been very en-couraging. Many differentoxidation agents were consid-ered, but sodium carbonate andcalcium carbonate were selectedbecause of desirable solubilityand stability properties. Theyare also essentially nonhazard-ous materials. Tests with pluto-nium metal in molten salts haveshown them to be effective inconverting the metal to the diox-ide. Experiments with wastesalts have also produced pluto-nium dioxide as the only pluto-nium species. Thermodynamicmodeling predicts that the onlymaterials that will be left in thedistillation heel will be pluto-nium dioxide and graphite. Afinal calcination step can be in-corporated into the process toconvert the graphite to carbondioxide, leaving a plutonium-dioxide heel suitable for storage.

Future WorkPreliminary results withnonoptimized equipment havealready demonstrated that distil-lation can be used to purify so-dium chloride-potassiumchloride salts so that they con-tain plutonium in the range ofhundreds of ppm. Our ultimateobjective is to produce salts thatcan be disposed of as LLW. Bet-ter equipment is being designedto achieve this objective.

Pretreatment of Waste SaltsThe calculations in the previous section assumed that the waste in-cluded only chloride salt and plutonium dioxide. Actual waste saltsmay also contain plutonium metal and plutonium trichloride andmust be pretreated to convert all actinide species to oxides. Oxygensparging can convert plutonium trichloride to plutonium dioxide.However, because neither plutonium metal nor oxygen is solublein the chloride salts to a significant extent, using oxygen spargingto convert plutonium metal to plutonium dioxide is more difficult.Plutonium oxychloride is also generally a product of oxygensparging and, based on preliminary vapor pressure measurements,9may not be desirable. Alternate oxidation agents that are solublein the salt are being examined experimentally and modeledtheoretically.

ResultsWe are still in the early stages of distillation experiments and are us-ing equipment that was easily available but not necessarily designedfor our purposes. Nevertheless, early results have been extremelyencouraging. Greater than 95% of the salt was recovered from plu-tonium-rich fractions of sodium chloride-potassium chloride saltspretreated by oxygen sparging. The plutonium concentration wasreduced from approximately 50% to the range of hundreds of ppm.Although still significantly above the LLW cutoff, this concentrationis low enough to allow very significant savings in costs for WIPPdisposal of Rocky Flats salts. Experiments with “cold salts” in theglovebox indicate that the level of contamination in the distilledsalts may be the result of handling in the glove-box environment,not the distillation process itself. The distillation heels left in thecrucible consisted of loose powders that could be poured out.



65


The encouraging preliminaryresults and the calculationspredicting an extremely lowplutonium concentration ofthe distillate salt lead us toconclude that the ambitiousgoal of converting plutonium-bearing waste salts to low-levelwaste is achievable.✦

Cited References1. S. Dushman, Scientific Founda-tions of Vacuum Technique, 2nd ed.(John Wiley and Sons, New York,1962).

2. J. L. Barton and H. Bloom, J. Phys. Chem. 60, 1413 (1956).

3. E. E. Schrier and H.M. Clark, J Phys. Chem. 67, 1259 (1963).

4. O. L. Hilenbrand and N. D. Pot-ter, J. Phys. Chem. 67, 2231 (1963).

5. D. A. Schultz and A. W. Searcy,J. Phys. Chem. 67, 103 (1963).

6. R. Hultgren, P. D. Desai, D. T.Hawkins, M. Gleiser, K. K. Kelly,and D. D. Wagman, Selected Valuesof the Thermodynamic Properties of theElements (American Society for Met-als, Metals Park, Ohio, 1973), p. 402.

7. T. E. Phipps, G. W. Sears, R. L.Siefert, and O. C. Simpson, J. Phys. Chem. 18, 713 (1950).

8. R. J. Ackerman, R. L. Faircloth,and M. H. Rand, J. Phys. Chem. 70,3698 (1966).

9. M. A. Williamson, Los AlamosNational Laboratory, personal com-munication, June 1993.


Fig. 3. Plutonium-richfraction of an oxygen-sparged salt before vacuumdistillation.

Fig. 4. Plutonium-dioxide powder heel and purified condensate salt.

Calcium chloride salts constitute another major fraction of the stock-piled waste salts, and equipment is being obtained to carry out thedistillation process at the higher temperatures required for calciumchloride.

Experiments are also being carried out on “cold” salts to helpoptimize processing parameters such as time, temperature, andpressure. These experiments will aid condenser and other equip-ment design. Surrogate metal oxides will be mixed with salts tohelp determine the extent to which the salts can be purified in anoncontaminated environment. Experiments with carbonateoxidation agents will also continue.


Evaluating Existing Partitioning Agents for Processing Hanford High-Level Waste Solutions

S. Fredric MarshNuclear Materials Processing:Nitrate Systems

Zita V. Svitra andScott M. BowenSeparation and RadiochemistryChemical Science and TechnologyDivision

“Los Alamos NationalLaboratory has been askedby DOE to support theHanford Tank WasteRemediation Systemmission, which is to 'store,treat, and dispose of all tankwaste in a safe, cost-effec-tive, and environmentallysound manner.'”

IntroductionThe Hanford Reservation in Washington State incorporates 177underground storage tanks that contain over 165 million curies ofradioactivity in more than 65 million gallons of waste. Many differ-ent processes were used at Hanford during the half-century thesewastes were accumulated from nuclear materials processing fornational defense requirements. As a result, many different reagentsand waste streams were neutralized and combined in undergroundtanks, often with insufficient concern about the compatibility of thecomponents being stored. The resulting wastes are an extremelycomplicated and sometimes unstable mixture of sludges, saltcakes,slurries, and supernates. Adding significantly to the stored wasteproblem is the fact that 67 tanks are known, or are suspected, tohave leaked.

The U.S. Department of Energy (DOE) is committed to the environ-mental remediation of hazardous wastes stored at the Hanford Site.Los Alamos National Laboratory has been asked by DOE to supportthe Hanford Tank Waste Remediation System (TWRS) mission,which is to “store, treat, and dispose of all tank waste in a safe,cost-effective, and environmentally sound manner.” An essentialrequirement for achieving the TWRS mission is identifying suitablepartitioning agents and technologies to accomplish this challenginggoal.

Experimental ProgramThe objective of our study was to evaluate the performance of alarge number of potentially useful absorber materials, includingmany not previously studied, in realistic waste simulants of acid-dissolved sludge and acidified supernate solutions found inHanford high-level waste (HLW) Tank 102-SY. Relatively fewabsorber materials have been evaluated with media approachingthe complexity of Hanford HLW solutions. Nor have other inves-tigators evaluated the distribution of such a large number of ele-ments onto so many absorbers.

We evaluated 60 commercially available or experimental absorbersfor partitioning high-level radioactive waste. We use the term“absorbers” to represent solid materials that remove ionic speciesfrom aqueous solutions, whether by ion exchange, chelation, iontrapping, molecular sieving, or other mechanisms. Absorbers inour study included cation and anion exchange resins, inorganicexchangers, composite absorbers, and a series of liquid extractantssorbed on porous support beads. We measured the distributionsof 14 elements onto these absorbers.


67

We also included some experi-mental absorbers to provideguidance for future researchactivities.

Waste SimulantsWe recognize that no wastesimulant will exactly representthe contents of an HLW tank.Nevertheless, we have used thebest information available inour effort to prepare realisticsimulants. Ultimately, the mostpromising absorbers must betested with actual waste.However, because actual wastesamples are expensive and diffi-cult to obtain, we suggest thatscreening of promising candi-date absorbers precede testswith actual Hanford tank waste.

Flow Sheet OptionsOur identification of suitablepartitioning agents was alsointended to help designers de-velop reliable flow sheets.The selection of agents will,of course, depend on the par-ticular objectives of the flowsheet. Although our study ini-tially was intended to supportthe flow sheet development pro-gram for Hanford Tank 102-SY,1

many of our experimental find-ings are applicable to other flowsheet proposals, such as the so-called “Clean Option.”2 Deci-sions about whether the targetedelements should be recoveredindividually, or grouped for sub-sequent disposal, are properlyleft to the flow sheet designers.Whichever flow sheet isadopted, we expect that ourexperimental data will help thedesigner anticipate the perfor-mance of existing partitioningagents.

The selected elements, whichrepresent fission products (Ce,Cs, Sr, Tc, and Y), actinides (U,Pu, and Am), and matrix ele-ments (Cr, Co, Fe, Mn, Zn, andZr), were traced by radionu-clides and assayed by gammaspectrometry. Distributioncoefficients for each of the 1680element/absorber/solutioncombinations were measuredfor dynamic contact periods of30 minutes, 2 hours, and 6 hoursto provide sorption kineticsinformation for the specifiedelements from these complexmedia. We measured more than5000 distribution coefficients.

AbsorbersMany organic and inorganicabsorbers are more resistantto chemical attack and radiationdamage than are most liquidextractants. Moreover, solidabsorbers can readily be usedin compact, low-cost columnsthat provide rapid multistageseparations. A typical operationcycle for a solid absorber col-umn consists of loading the ab-sorber and elution of the sorbedelement(s), followed by regen-eration and reuse of the ab-sorber. At times, the targetedelements are sorbed so stronglythat elution is difficult; in suchcases, the loaded absorber be-comes a candidate for the finalwaste form.

Although most of the absorberswe evaluated are cation ex-changers, we included some an-ion exchange resins and a seriesof liquid extractants (sorbed onporous carbon beads) to obtainmore comprehensive data, aswell as to supply convenientreference points to other studies.



ResultsOnly a brief summary of ourresults is presented here.A detailed description and allexperimental data have beenpublished elsewhere.

3

Behavior for many absorber/element combinations was sub-stantially different from whatwould be expected, based onpublished measurements fromrelatively clean solutions; thisfinding demonstrates theimportance of using realisticsimulants. In some cases, wefound that inexpensive commer-cial partitioning agents outper-form specialty products whosecosts are orders of magnitudehigher.

In addition to identifyingpromising existing partitioningagents, we determined whichpartitioning agents and tech-nologies, including some ofthose that have been designatedas the Hanford baseline tech-nologies, are marginal or inad-equate. Table I summarizeshow well each of the 14 ele-ments is sorbed, from the twoacidic solutions, by showing thenumber of effective absorbers ineach of the six distribution coef-ficient (Kd) categories. Whereacceptable absorbers have beenidentified, we recommend thatthese absorbers be tested withother simulated tank composi-tions and with actual waste.Where acceptable absorbershave not been identified, we rec-ommend an accelerated effort toidentify or develop satisfactorypartitioning agents.


Kd Kd Kd Kd Kd Kd KdMedium Element >1000 300-1000 100-300 20-100 10-20 5-10

Acid- Ce 0 0 0 5 1 1Dissolved Cs 7 0 0 11 1 4Sludge Sr 0 0 0 0 0 0

Tc 0 0 1 11 2 5Y 0 0 0 0 0 0Cr 0 0 0 0 0 0Co 0 0 0 0 1 0Fe 0 0 0 0 0 0Mn 0 0 0 0 0 0Zn 0 0 0 0 0 0Zr 0 1 3 10 4 3U 0 0 0 6 8 5Pu 2 5 6 2 8 2Am 0 0 0 0 0 6

Acidified Ce 1 3 3 10 3 6Supernate Cs 8 1 5 2 4 2

Sr 0 0 0 0 3 4Tc 0 7 7 9 3 1Y 1 0 1 9 0 5Cr 0 0 0 2 2 5Co 0 0 0 1 3 4Fe 0 0 0 0 0 3Mn 0 0 0 0 1 5Zn 0 0 0 5 3 2Zr 0 0 0 1 4 3U 0 1 4 15 8 8Pu 0 0 0 0 0 1Am 0 3 2 13 1 5

Table I. Number of Identified Absorbers in Various Kd Categories for Each of14 Elements, from Two Solutions Simulating Hanford HLW Tank 102-SY

Elements that show good-to-excellent sorption from the simulantsolution of acid-dissolved sludge are cerium, cesium, technetium,zirconium, uranium, and plutonium (Table II). Elements that showlow sorption from this solution are strontium, yttrium, chromium,cobalt, iron, manganese, zinc, and americium.



69

simulant solutions selected forthis study, some of the mostpromising absorbers identifiedin this study may have applica-tions at TA-55. This is especiallylikely for those absorbers andextractants that sorb americiumstrongly. Testing of likely ab-sorbers with TA-55 acidic andalkaline waste is planned.

Future StudiesOur screening study has met itsobjective of identifying specificabsorbers that appear capableof partitioning targeted elementsfrom Hanford HLW Tank 102-SY. The best absorbers from thisstudy and perhaps a few prom-ising new materials should beevaluated with simulated com-positions that represent other

Hanford tanks, especially thoseon the safety watch-list. The 14elements whose distributionshave been measured should becarefully examined to determineif any should be dropped or ifadditional elements should beincluded in future measure-ments. Finally, radioactive trac-ers, shown to provide a rapid,reliable, and inexpensive wayto obtain large amounts of distri-bution data, should be used insuch studies.

The effects of radiation on themost promising absorbers fromour present and future studiesshould be determined, whensuch information is not alreadyavailable, 4 before partitioningagents are selected or before anylarge-scale recovery flow sheet isselected.

Elements that show good-to-excellent sorption from thesimulant solution of acidifiedsupernate are cerium, cesium,technetium, yttrium, zinc, ura-nium, and americium (Table III).Elements that show poor-to-marginal sorption from this so-lution are strontium, chromium,cobalt, iron, manganese, zirco-nium, and plutonium.

Applications to Los AlamosPlutonium Facility WasteThe Los Alamos PlutoniumFacility at TA-55 has assigned ahigh priority to waste minimiza-tion. Because the acidic wastestreams generated at TA-55 arein some ways similar to the



Table II. Absorbers* Whose Measured Distribution Coefficients Exceed 20for the Specified Elements, from Simulated Acid-Dissolved Sludge Solution

Medium Element Absorbers

Acid- Ce JSK-1, JSK-2, JSK-3, PuroliteTM A-520-E, IonacTM SR-3DissolvedSludge Cs SNL/CSTs, AMP-PAN, M315-PAN, NiFC-PAN

Tc ReillexTM HPQ, IonacTM SR-3, PuroliteTM A-520-ESybronTM (Et)3N, LANL TiO2, LANL ZrO2, LANL Nb2O5

Zr NaY-PAN, SNL/CSTs, DuracilTM 230, TiO-PAN, IonsivTM TIE-96,M315-PAN, DuoliteTM CS-100, CyanexTM 272

U TRU-SpecTM, Sybron (Et)3N, IonacTM SR-3, PuroliteTM A-520-E

Pu JSK-3, JSK-2, JSK-1, ReillexTM HPQ, TRU-SpecTM, Sybron SR-3,PuroliteTM A-520-E, CyanexTM 923

* Detailed descriptions of these absorbers, whose trade names are used here, have been published elsewhere (see Reference 3).


Medium Element Absorbers

Acidified Ce CyanexTM 923, TRU-SpecTM, SNL/HTO, CMPO-DIPB, AmberlystTM 15,Supernate DHDECMP, MgO-PAN, DiphonixTM, SNL/CST111, AmberlystTM XN-1010

Cs SNL/CSTs, NiCF-PAN, AMP-PAN, M315-PAN, DurasilTM 230,SRS resorcinol, IonsivTM TIE-96

Tc Sybron (Pr)3N, CyanexTM 923, AliquatTM 336, Sybron (Et)3N,

ReillexTMHPQ, PuroliteTM A-520-E, IonacTM SR-6, IonacTM SR-3,TRU-SpecTM, LIXTM-26, DHDECMP, CMPO-DIPB, LANL TiO2,LANL ZrO2, LANL Nb2O5

Y CyanexTM 923, TRU-SpecTM, AmberlystTM 15, MgO-PAN,CMPO-DIPB, AmberlystTM XN-1010

Zn NiFC-PAN, DuoliteTM C-467, NaTiO-PAN

U CyanexTM 923, DiphonixTM, TRU-SpecTM, TiO-PAN, SNL/HTO,CMPO-DIPB, SNL/CST35, CyanexTM 272, DuoliteTM C-467,

AmberlystTM 15

Am CyanexTM 923, TRU-SpecTM, SNL/HTO, CMPO-DIPB, AmberlystTM 15,SNL/CST111, AMP-PAN, Bone char, AmberlystTM XN-1010,

LIXTM-1010

Table III. Absorbers* Whose Measured Distribution Coefficients Exceed 50for the Specified Elements, from Simulated Acidified Supernate Solution

M. Bowen, “Distributions of 14 Ele-ments on 60 Selected Absorbersfrom Two Solutions (Acidic Dis-solved Sludge and AlkalineSupernate) Simulating HanfordHLW Tank 102-SY,” Los AlamosNational Laboratory report LA-12654 (October 1993).

4. S. F. Marsh and K. K. S. Pillay,“Effects of Ionizing Radiation onModern Ion Exchange Materials,”Los Alamos National Laboratoryreport LA-12655 (October 1993).

ConclusionsOur experimental data indicatethat many existing partitioningagents may be suitable for HLWtank remediation. If reliablepartitioning agents can be iden-tified, Hanford HLW tank pro-cessing might begin and reachcompletion sooner, and at alower cost, than would other-wise be possible. We thereforefeel that the findings of ourstudy could have a major benefi-cial impact on environmentalremediation efforts at Hanfordand elsewhere within the DOEcomplex.✦

Cited References1. S. L. Yarbro et al., “Status of Tank102-SY Remediation Flowsheet De-velopment at LANL,” Los AlamosNational Laboratory report LA-UR-93-1725 (April 1993).

2. J. L. Straalsund et al. “Clean Op-tion: An Alternative Strategy forHanford Tank Waste Remediation:Volume 1. Overview,” Pacific North-west Laboratory report PNL-8388Vol. 1 (December 1992).

3. S. F. Marsh, Z. V. Svitra, and S.

* Detailed descriptions of these absorbers, whose trade names are used here, have been published elsewhere (see Reference 3).


Evaluating Existing Patitioning Agents for Processing Hanford High-Level Waste Solutions

71


Gas Pycnometry for Density Determination of Plutonium Parts

J. David Olivas andS. Dale SoderquistPlutonium Metallurgy

Henry W. Randolph,Susan L. Collins,Darren Verebelyi,William J. Randall,and Gregory D. TeeseEquipment Engineering SectionWestinghouse Savannah RiverCompany

IntroductionTraditionally, determination of plutonium component density hasbeen accomplished by measuring the apparent weight loss whenthe object is immersed in a liquid of known density (Archimedes’principle). Liquids that are compatible with plutonium, and thathave been successfully used in the past, include freon andmonobromobenzene. However, these fluids have undesirable char-acteristics that present health and environmental hazards. Further-more, the spent fluid stream is environmentally undesirable. Gaspycnometry eliminates the liquid waste as well as the gaseous emis-sions from volatiles. The focus of current gas pycnometry researchand development is to prove that this technique can reliably deter-mine the density of plutonium subscale shapes to the desired accu-racy. If this goal is met, gas pycnometry can be expanded toaddress all plutonium density determination needs.

BackgroundThe accurate determination of part density is an important param-eter in plutonium fabrication. Densities of solids are routinely de-termined by the hydrostatic technique of measuring the apparentweight loss when the solid is submerged in a liquid of knowndensity.1 When the sample volume, determined from Archimedes’principle, is combined with the true weight of the sample, its den-sity is readily calculated. However, the hydrostatic technique issubject to error if the sample surface has cracks that are not com-pletely filled by the liquid. Similarly, trapped air bubbles can intro-duce errors. Furthermore, the thermal effects of heat-generatingmaterials such as plutonium can induce thermal currents or raisethe temperature of the fluid and thus affect the accuracy of the re-sults. If the fluid chemically reacts with the sample, or adsorbs onthe surface of the sample, the sample will be physically altered, andresults will be inaccurate.

Accuracy of the hydrostatic technique for determining plutoniumpart density thus depends, in large part, on the choice of fluid.Freon and monobromobenzene, the traditional fluids of choice forplutonium, are less attractive now because of emerging environ-ment, safety, and health regulations. Freon production is being cur-tailed by the Montreal Protocol, whereas monobromobenzene is asuspect carcinogen. Searches for other liquids were conducted, andsome were found that provided adequate results. Unfortunately,these liquids required the use of a solvent to remove them from theplutonium surface when density determination was complete. Theincreasing regulation of liquid waste disposal prompted efforts toidentify a method of density determination that is not dependentupon the hydrostatic technique.

“The focus of current gaspycnometry research anddevelopment is to provethat this technique canreliably determine thedensity of plutoniumsubscale shapes to thedesired accuracy. If thisgoal is met, gas pycnometrycan be expanded to addressall plutonium densitydetermination needs.”


Los Alamos researchers entered into a technology developmentpartnership with equipment development specialists from the Sa-vannah River Technology Center (SRTC), a division of theWestinghouse Savannah River Company. The goal of this partner-ship was to design, build, and test a prototype system to demon-strate the application of gas pycnometry on a subscale plutoniumshape. Requirements developed for the gas pycnometer systemwere as follows:

• determines density within ± 0.05%,• does not leave residues or contaminants on part,• accommodates self-heating of plutonium,• operates in a glove-box environment, and• completes measurement cycle in less than two hours.

Prototype Gas Pycnometer SystemA schematic of the device developed by SRTC is shown in Fig. 1.The sample chamber is made of a solid block of copper to providethe required heat transfer characteristics for the temperature controlsystem. The designers selected a simple-to-machine, hemisphericalshell for the sample part. The chamber closely matches the shapeof the sample so that the sample will occupy a significant portionof the available volume. The chamber volume should be within 50%of the volume of the part to allow a sufficiently large change in gas

Bellows

GasInjectorActuator

CopperSampleChamber

Bellows

Fig. 1. Schematic of prototype gas pycnometer.



Gas pycnometry is used success-fully in industries that are ill-suited for hydrostatic densitydetermination. These industriestypically process powders, suchas metal powders, ceramic pow-ders, and dry cement. To usethis technique, a volume of gasis allowed to expand in a cali-brated sample chamber, and thepressure difference caused bythe sample volume is calculated.Different implementations arepossible, but all rely upon thefundamental gas law. Equation1 is the fundamental gas law atconstant temperature and solv-ing for the new volume V2yields Equation 2:

P1V1 = P2V2 (1)

and

V2=V1(P1/P2). (2)

The application of gaspycnometry to density determi-nation of plutonium parts ismore complicated than normalindustrial applications, prima-rily because plutonium is self-heating and will increase thetemperature of the gas in thepycnometer. The fundamentalgas law at nonconstant tempera-ture and its solution for volumeV2 are shown in Equations 3and 4:

(P1V1)/T1=(P2V2)/T2 (3)

and

V2=V1(P1/P2)(T2/T1) . (4)

In addition, the precision de-sired for plutonium fabricationis much higher than standardindustrial practice. Consideringthese factors, we concluded thata system designed specificallyfor use with plutonium wouldbe required.

73

The set point is selected by thecomputer to ensure that heat isalways being removed. In addi-tion to maintaining the device atthe set point temperature, thethermoelectric coolers have suf-ficient capacity to remove theheat generated by a plutoniumsample.

An actuator on the side of thedevice injects a known, repeat-able amount of gas into thechamber. The pressure increasecaused by the gas injection isused to calculate the chambervolume from the fundamentalgas law. The difference betweenpressures in the chamber withthe sample present and absent isused to calculate the sample vol-ume. Pressure is measured witha highly accurate quartz pres-sure transducer. The computerreads the transducer for twentymeasurements before releasingthe actuator to relieve gas injec-tion pressure, and then twentylow pressure readings are taken.Readings are taken multipletimes to establish stability andgive statistical significance tothe values used in subsequentcalculations.

A computer running customdata collection and device con-trol software controls the entiresystem. The computer does notallow a measurement cycle tobegin if temperature stability isnot achieved. Once a measure-ment has been initiated, no userinteraction is required; thus, op-erator variability is not a factorin the measurement process.

An aluminum hemisphericalshell was fabricated for initialout-of-glove-box testing of thedevice. In addition, a stainless-steel shell encircled by resistance

heating wire was constructed tosimulate the self-heating proper-ties of plutonium. A plutoniumshell has been designed and isto be fabricated at Los Alamos.

Savannah River cold-tested thegas pycnometer with both thealuminum and stainless-steelshells. The device was thentransported to Los Alamoswhere it was cold-tested againwith the aluminum shell.

Aluminum Shell Test Results atSavannah RiverThe aluminum shell was usedfor most of the Savannah Riverexperiments. Its true volumewas determined by the hydro-static technique, using water asthe working fluid. The immer-sion was performed in a vacuumso that air bubbles could notform and bias the results. Themean of the results of a numberof trials was used as the best es-timate of the true part volume.The standard deviation of thehydrostatic measurements was0.08%.

A number of trials were thenperformed using the aluminumshell in the gas pycnometer.The repeatability of the resultswas superior to the hydrostatictechnique (0.01% at 1 std dev).However, an offset of the meanwas noted. Analysis of the re-sults indicated that the sweptvolume of the actuator, andhence the volume of the gasinjection, was different than thetheoretical value because of me-chanical tolerances in mountingthe actuator. The correct valuefor the actuator volume wasthen calculated and used infuture measurements.

pressure and thus provide rea-sonable accuracy. Other notablefeatures of the device includethe rubber bellows in the bottomportion for opening and closingthe chamber and the side-mounted actuator for injectingthe gas charge.

To achieve the ± 0.05% repeat-ability specification for partdensity, variations from onemeasurement cycle to the nexthad to be carefully controlled.An analysis of measurementuncertainties indicated that oneof the major design challengeswas closure of the sample cham-ber to achieve a highly repeat-able volume. A rubber bellowsactuator was selected to lowerand raise the male portion of thechamber in order to open andclose the chamber. The pressurein the bellows is controlled towithin 0.05 psi to assure consis-tent compression of the O-ringseal. Also, the bellows accom-modates angular misalignment,and this inherent property en-sures that pressure is applieduniformly around the O-ringcircumference.

Temperature is controlledthrough a feedback controlloop. Thermistors are used tomeasure the temperature within± 0.001 K. The thermistor valuesare read by the control com-puter, which controls the heatremoval rate by adjusting thevoltage output to the thermo-electric coolers. Heat must beconstantly removed from the de-vice to maintain the temperaturewithin ± 0.01 K of the set point.




75.1

75.15

75.2

75.25

75.3

75.35

75.4

0 5 10 15 20

Aluminum Hemisphere Data from LANL Trials

Volu

me

(in

Cu

bic

Cen

tim

eter

s) Upper Goal

Lower Goal

95% ConfidenceBand

Trial Number

Stainless-Steel Shell TestResults at Savannah RiverThe stainless-steel shell had aseries of concentric groovesmachined into the outer surface.Resistance heating wire waswound into these grooves andconnected to a power supply.By varying the supply voltage,the power of the heater was ad-justed to approximate the powergenerated by a plutonium shell.The power dissipation duringtrials with the stainless-steelshell was 2.5 watts.

As anticipated, trials conductedon the heated stainless-steelshell showed a loss in precision.The loss is attributed to unevenheat generation, which leads tovariations in the temperature ofthe gas in the annular space.However, the observed precisionof 0.03% is still better than thegoal of 0.05%. Uniform heatgeneration from plutoniumshould reduce the variabilitysomewhat.

Aluminum Shell Test Resultsat Los AlamosAfter transport of the gas pyc-nometer to Los Alamos, testingwith the aluminum shell wasrepeated to confirm that thedevice was not damaged intransit. Repeatability within0.01% was achieved, confirmingproper operation of the deviceand reconfirming the resultsfrom Savannah River. Theresults of this testing arepresented in Fig. 2.

Future DirectionsPreparations are in progress forinstalling the gas pycnometer ina glove box in the Los AlamosPlutonium Facility. After instal-lation, the aluminum shapewill again be measured withthe gas pycnometer and then

Fig. 2. Los Alamos aluminum shell data compared to experimental goals.

measured with the hydrostaticmonobromobenzene method.No further testing is plannedwith the heated stainless-steel shell; instead, the pluto-nium shell will be used.The plutonium shell will bemeasured with both the gaspycnometry and hydrostaticmonobromobenzene methodsto determine the success of thepycnometer system.

The pycnometry method is in-herently less flexible than thehydrostatic technique becauseof the geometric constraints ofthe pycnometer sample cham-ber. Near-term efforts includefabricating a pycnometer systemthat will accommodate specificpart geometries for Los Alamosmissions. A longer-term effortwill examine alternative meth-ods of increasing robustness,such as replaceable chamberinserts or premeasured gasvolumes at higher pressures.Eventually, gas pycnometrywill replace hydrostatic densitydeterminations for plutonium.

SummaryLos Alamos scientists, in coop-eration with their SRTC part-ners, are evaluating gaspycnometry for determiningdensity of plutonium parts.This replacement for the currenthydrostatic immersion tech-nique will eliminate a liquidwaste stream and support thevision of an environmentallybenign plutonium facility.Initial testing of gas pycnometrywith surrogate materials indi-cates the technique can provideresults that are superior to thosefrom existing methods. Densityof aluminum hemisphericalshells can be determined to ±0.01% of the actual density. Test-ing with plutonium shells willcommence in the near future.✦

Cited Reference1. F. W. Dykes and J. E. Rein, “Ordi-nary Weighing and Density Deter-mination,” Progress in NuclearEnergy X, 11-34 (1970).



75


Plutonium Dry Machining

Mark R. MillerPlutonium Metallurgy

“Reducing or eliminatinghazardous wastes fromplutonium machiningprocesses at Los AlamosNational Laboratory iscritical to ensure continuedoperations under currentand increasing legislativedemands.”

IntroductionReducing or eliminating hazardous wastes from plutonium machin-ing processes at Los Alamos National Laboratory is critical to en-sure continued operations under current and increasing legislativedemands. In the past, Los Alamos machining processes used an ar-gon and freon mist system to lubricate and cool the plutonium partsand to act as a fire inhibitor (plutonium is pyrophoric). Current U.S.legislation requires the elimination of all chloro-fluorocarbons, suchas freon, by 1996; therefore, switching to plutonium machiningmethods that are environmentally friendly is crucial. Dry machin-ing is one possibility because it does not require the use of coolantsor oils, and, thus, mixed waste byproducts can be minimized oreliminated. The impact of this method is realized downstream aswell, because the material recovery processes receive cleaner feedmaterial.

Dry machining tests at Los Alamos on various plutonium parts,which included features similar to weapon components, showedthat plutonium can easily be dry machined, practically eliminatingthe need for cutting fluids. Results indicated a significant reductionin the use of oils and solvents for research and development activi-ties as well as for production processes. This reduction and eventualelimination of hazardous solvents and oils for plutonium machiningprocesses will complement activities at the future plutonium pro-duction facility, Complex 21.

BackgroundMetal machining has traditionally used cutting fluids to lubricate orcool parts at the cutting tool interface during the material removalprocess. These fluids have a number of benefits, such as reducingfriction and wear (thus extending tool life and improving surfacefinish), cooling of the cutting zone, and removing metal chips fromthe immediate work area. The fluids are generally water-based oil,organic oils, or solvents. However, plutonium machining has beenassociated only with compatible organic oils or solvents, becausewater acts as a neutron reflector that decreases the amount of pluto-nium needed to sustain a nuclear chain reaction.

Cutting fluids can be used as lubricants or coolants depending ontheir applications. High-speed machining requires coolants to re-duce heat that occurs from increased cutting speeds, whereasslower-speed machining requires lubricants to reduce materialbuildup on tooling and to improve surface finish. Past plutoniummachining processes have used cutting fluids for both purposes,to lubricate and to cool parts.




The plutonium shapes typically machined require that the cuttingtool begin at the spindle center line and proceed outward to somespecified diameter, or the reverse of this process. The operationsare analogous to a facing operation performed from the center lineoutward. A direct relationship exists between the cutting speed andthe radius of the part; as the radius (distance from the center line)increases during machining, the part surface speed increasesin relation to the cutting tool.1 Therefore, close to the spindle centerline, a rotating plutonium part has very low surface speed and thecutting process acts as a low-speed machining operation. As thetool progresses outward, the surface speed of the part increases andthe cutting process behaves more like a high-speed operation. Thisdirect relationship between the cutting speed and the part radius isexpressed as

Vc = πr × ω ,

where

VC = cutting speed (ft/min), r = radius (ft), and ω = rotational speed (rpm).

Plutonium is a material that must be processed in a controlledenvironment in order to prevent plutonium contamination, limitradiation exposure to personnel, and reduce the amount of oxygenpresent. Because plutonium is pyrophoric, heat created from pro-cessing must be regulated, along with oxygen, to prevent ignitingthe metal. Reducing oxygen content is also necessary to inhibitoxide growth (plutonium reacts with oxygen to create plutoniumoxide). Oxygen reduction within a glove box requires purging itwith either nitrogen or argon.

Los Alamos and the Rocky Flats Plant each have used a differentapproach for cooling and lubricating the plutonium parts. Produc-tion machining methods at Rocky Flats used oil as a cutting fluidthat was applied by flood cooling. This method served not onlyto cool and lubricate the parts but also limited the amount of oxy-gen in contact with the material at the tool and part interface.However, the oil, once in contact with plutonium, is considered amixed waste and was difficult to dispose of properly. Additionally,the oil had to be cleaned from the surface of the parts, chips, andequipment with large amounts of carbon tetrachloride, approxi-mately 7800 gallons per year. This solvent, when contaminatedwith plutonium, is considered a mixed hazardous waste and re-quires very strict disposal procedures. Research and developmentplutonium machining methods at Los Alamos used argon to propela freon mist onto the tool and part interface for the same reasons

as at Rocky Flats. Freon usage atLos Alamos was approximately50 liters per year. Because freonis a chloro-fluorocarbon, its usewill soon be eliminated in theUnited States. Clearly, alterna-tives to these environmentallyharmful practices at Rocky Flatsand Los Alamos were desper-ately needed.

Rocky Flats investigated alterna-tive solvents for carbon tetra-chloride while Los Alamossearched for methods to elimi-nate freon; however, Rocky Flatswas shut down before an alter-native could be found. The re-search performed at Los Alamosrevealed that by reducing theoxygen content low enough inthe glove box, plutonium couldbe dry machined, reducing theneed for cutting fluids. Cur-rently, this process is appliedin approximately 85% of allmachining operations atLos Alamos. The remainingoperations are carried out witha light, thin coat of oil appliedto the surface of the material forfinish cuts.

MethodologyTests at Los Alamos were per-formed as an engineering studyto investigate the feasibility andpracticability of plutonium drymachining for research anddevelopment purposes as wellas for production processes.Experiments included upperoxygen limit and plutoniummachining tests. In addition,the formation of nitrogen andplutonium compounds were in-vestigated in the currentliterature.

77



At this oxygen level the burningof the plutonium was concen-trated in the turnings (metalchips) as they fell from the part.As the turnings burned, theyquickly changed to Pu2O3 as theheat caused by the friction of thecutting tool in combination withthe 4.5% oxygen created thecritical mix of plutonium, heat,and oxygen needed for combus-tion. Experience with pluto-nium machining and occasional“fires” has shown that fires areindeed associated with chipsonly. This concentration of heatenergy, approximately 80%,occurs because of the frictionbetween the cutting tool rakeface and the chip, typically re-sulting in tool crater wear.Because maximum temperaturesare located away from the cut-ting tool edge,2 the plutoniumpart does not receive the fric-tional heat energy that is carriedaway with the chips. The ex-perimental results, in combina-tion with plutonium chip fireexperience, were used to estab-lish a maximum oxygen levelof 3% for dry machining ofplutonium parts.

Machining Tests

Results from the machining testsshowed no detectable differ-ences in quality between partsmachined without cutting fluidsand those machined with fluids.The dry-machined parts wereinspected on a AA-gage, a rotarycontour inspection machine, thatproduced inspection data con-sistent with data from parts ma-chined with cutting fluids.No quantitative surface finishdata could be gathered becauseno equipment exists in glove

Establishment of an upper oxy-gen limit for dry machining wascritical to the prevention of aplutonium fire. Two methodswere used to test the oxygenlimit at which plutonium wouldignite under various machiningconditions. The first machiningtests were performed on an en-gine lathe. A series of cuts weretaken on a 0.75-inch, delta-stabi-lized plutonium rod while theoxygen content in the glove boxwas increased. Another seriesof tests were performed in a fur-nace to verify the temperature aswell as oxygen content requiredfor plutonium combustion.In both approaches, only smallamounts of plutonium wereused in order to prevent a largefire from occurring.

The machining investigationswere conducted on variousphases of plutonium, includingdelta-stabilized material. Thetests concentrated on turning(the shaping process) and wereperformed on computerized nu-merical control T-base lathes.Various shapes were machinedincluding component parts thatwere designed for, and tested at,the Nevada Test Site. These full-size parts were of the type thatwould be manufactured forweapons production.

ResultsUpper Oxygen Limit Tests

Test results for the safe upperoxygen limit during a pluto-nium machining operationshowed that the glove-boxoxygen level must increase toapproximately 4.5% before plu-tonium ignition could occur.

boxes at Los Alamos to performthis inspection. However, experi-enced inspectors estimated thesurface finish to be as good asparts machined with cuttingfluids.

Test parts, along with productionparts, showed that, when facingoperations are required, dry ma-chining plutonium can gall (ortear) the material at the centerline of the rotating part. Gener-ally, the size of this defect wasapproximately 0.050 inches wideby 0.000080 inches deep, and thisdefect appeared in approxi-mately 30% of all parts that re-quired this type of operation.Although these plutonium partspass inspection criteria, they donot meet the approval of theLos Alamos user. The physicistsand engineers who design thecomponents for testing at theNevada Test Site require thesecomponents to surpassweapons-grade specifications,physically and visually. Thisanomaly may cause a part to berejected even though the partmeets weapons-grade specifica-tions. However, a small amountof light oil swiped on the surfaceeliminated any galling.

The positive test results for ma-chining convinced Los Alamos toadopt dry machining as the alter-native to producing parts with anargon/freon mist system for allNevada Test Site componentsand for related Los Alamos ma-chining activities. About 85% ofall plutonium components aredry machined. The remaining15% require a light coat of oil toeliminate the potential for micro-inch tearing during the final cut.




The light coat of oil is brushedon the part just prior to perform-ing the finish cut. The oil-con-taminated chips, kept separatefrom the dry-machined chips,are cleaned with a small amountof freon. The oil-free chips arethen melt-consolidated undercalcium chloride to produce aplutonium button ready for fur-ther processing.

Nitride Investigations

Exposing plutonium to a nitro-gen atmosphere at elevated tem-peratures for extended periodsof time can create a compoundcalled plutonium nitride. Pluto-nium nitride is extremely un-stable in air and readily forms ahydrated oxide. Plutonium ox-ide can deteriorate the pluto-nium metal surface, reducedensity, and increase partweight.

However, the temperatures andtime required to form even smallamounts of a nitride compoundare quite high. Experimental re-sults have shown that 3% nitrideis formed at 625 °C after 16hours of exposure to nitrogen.3Estimates of temperatures atthe tool part interface are signi-ficantly lower, approximately300 – 400 °C, while cutting toolcontact times are approximatelymicroseconds.2,4 Therefore,even at 400 °C the creation ofany significant amount of pluto-nium nitride is highly improb-able because the nitride does nothave adequate time to form.

The 20 cubic centimeters of oilnow introduced annually intothe machining process easilymeets current regulations.

Flood cooling techniques atRocky Flats created 2500 gallonsof mixed waste oils each yearand required the use of approxi-mately 7800 gallons of carbontetrachloride to clean the oil-contaminated chips, machinetools, and parts. Application ofthe dry machining techniquesoutlined in this report (adjustedfor the number of units ma-chined at Rocky Flats) wouldentirely eliminate hazardous(carcinogenic) carbon tetrachlo-ride use, reduce the amount ofmixed waste oils to approxi-mately 1.25 gallons per year, andproduce approximately 50 gal-lons of waste freon each year.Obviously, dry machining tech-niques will complement the fu-ture plutonium productionprocesses.

Ongoing ResearchThe current trend in environ-mental legislation clearly showsthat the small amount of oil andfreon required for the dry ma-chining process should be elimi-nated. Plutonium machiningresearch and development ef-forts are addressing this issue.Efforts include the investigationof diamond turning techniquesto increase the cutting tool edgequality and to reduce friction onthe rake face. These performanceimprovements should produce a

Additionally, no rapid formationof oxide has been observed ondry-machined plutonium.Observations of dry-machinedparts have taken place over timeperiods of up to one year. Cer-tainly, if significant amounts ofnitride were formed during drymachining a yellow-green oxidegrowth would quickly attack theouter surface and would bereadily visible. The empiricaland physical evidence clearlyshows that the formation of plu-tonium nitride during the dry-machining process isinsignificant.

Waste Minimization and OtherBenefitsBenefits from dry machining ofplutonium parts include im-proved part quality, a signifi-cant decrease in chloro-fluorocarbon releases, and sig-nificant mixed waste reduction.

Part quality at Los Alamos wasenhanced by eliminating thefreon mist that produced a local-ized cooling effect. This local-ized cooling physically reducedthe part size in the immediatearea of coolant application dur-ing machining and thus pro-duced out-of-toleranceconditions.

Currently, plutonium dry ma-chining activities at Los Alamosuse approximately one liter offreon annually, as contrastedwith the 50 liters required forthe argon/freon mist system.

79



better surface finish as the toolapproaches the center line of arotating part. Changing spindlespeeds and cutting tool feedrates should improve the surfacefinish in a slow-speed cuttingoperation. Alternative cuttingfluids are also being explored,and this search has lead to a sol-vent called SF2I, developed bythe 3-M Company. This solvent,originally tested at the AtomicWeapons Establishment inAldermasten, England, showspromising results for plutoniummachining operations. It will befurther tested at Los Alamos forapplication as a coolant for ma-chining operations, such as saw-ing, drilling, and milling.♦

Cited References1. E. Oberg, F. D. Jones, andH.L. Horton, Machinery’s Handbook,22nd ed. (Industrial Press, Inc.,New York, 1987), p. 1765.

2. S. Kalpakjian, ManufacturingProcesses for Engineering Materials(Addison-Wesley, Reading, Massa-chusetts, 1991), pp. 499-500.

3. F. Brown, H. M. Ockenden, andG. A. Welch, “The Preparation andProperties of Some PlutoniumCompounds Part II,” J. Chem. Soc.1955, 4196-4201 (1955).

4. F. Kreith, Principles of HeatTransfer (Harper and Row, NewYork, 1973), p. 81.




“In this report, an assess-ment of issues relevant toplutonium storage is pre-sented in order to provide atechnical basis for address-ing complex interfaces withpolitical and economicissues and for identifyingsuitable storage optionsfor excess plutonium.”

IntroductionThe retirement of large numbers of nuclear weapons necessitates themanagement of an unprecedented quantity of excess plutonium. Sur-plus materials from production and reprocessing operations will addto the amount of concentrated plutonium that must be stored on aninterim basis until a decision regarding ultimate disposition is made.Traditionally, the weapons complex faced a scarcity of plutonium.Thus, the future course is unfamiliar. An evaluation of technical is-sues is needed to develop a basis for storage decisions. Identificationof acceptable storage options is a prerequisite to the definition of stan-dards, procedures, and facilities for storage.

The selection of suitable storage options is a complex task impactedby political and economic concerns as well as technical issues.Many factors influence decisions on plutonium storage; however,emphasis is often placed on a narrow set of issues, such as per-ceived proliferation resistance or some specific hazard. A compre-hensive examination should include topics such as the potential forremediation of hazards, the maintenance of flexibility for ultimatedisposition, and the cost of implementation.

A consideration of cost must extend beyond the initial expendituresfor research, development, and capital investment in facilities toother factors that contribute to the total expense of the operation.A thorough assessment must include not only the cost of addressingradiation exposure and contamination control but also the cost ofother environmental, safety, and health (ES&H) issues. For example,disposal of radioactive waste constitutes a major and increasinglylarger fraction of the total operating cost of a plutonium facility.The amount of waste that must be considered includes that pro-duced during normal facility operation as well as that generatedduring facility modification and decommissioning.

In this report, an assessment of issues relevant to plutonium storageis presented in order to provide a technical basis for addressing com-plex interfaces with political and economic issues and for identify-ing suitable storage options for excess plutonium. An effort is madeto summarize the merits of alternative approaches and to establish aqualitative ranking of options for an interim storage period of up toone hundred years. This work focuses on plutonium extracted fromweapons. It is important to note that this source represents the mi-nority of available plutonium. Considerable quantities of reactor-grade plutonium are found throughout the world and represent asignificant environmental and proliferation risk. Narrow solutionsthat only address material extracted from weapons may fail in thelarger context of a worldwide plutonium economy.

John M. Haschkeand Joseph C. MartzPlutonium Metallurgy

81

form and then reconfigured forpreparing material for reuse orfinal disposition. Finally, a real-istic assessment must includedecontamination and decom-missioning of interim processesas well as installation, operation,and decommissioning of anysubsequent facilities. The im-pact of such operations on cost,including waste generation dur-ing the entire cycle, must beweighed against that of storingmaterial in its existing form untilspecifications for final disposi-tion are defined.

The issue of special nuclear ma-terial (SNM) control also meritsexamination. In addition to ex-isting standards for accountabil-ity and verification of plutoniuminventory, increased transpar-ency requirements are antici-pated during an extendedstorage period. Certain optionsare more amenable to account-ability than others because thechemical form of the stored ma-terial is homogeneous. Whereasplutonium metal and com-pounds formed by gas-solid re-actions tend to be homogeneous,compounds prepared by con-densed-phase reactions or py-rolysis of precipitates arefrequently inhomogeneous andthus more difficult to inventoryand verify. Such assay and mea-surement is particularly difficultfor dilute plutonium forms, es-pecially those mixed with otherradioactive elements.

Technical IssuesConsideration of a broad rangeof issues is necessary for assess-ing plutonium storage options.A summary of the major issuesis included to provide a compre-hensive perspective.

Proliferation of nuclear weaponscapability by major nuclearstates, aspiring nations, orsubnational groups presents aserious threat to world orderand stability. The selection ofstorage options is complicatedby political pressures demand-ing that a technical remedy befound to prevent proliferation.The level of interdiction re-quired to prevent an existingnuclear state from rapidly recon-structing a nuclear arsenal fromintact pits differs from that re-quired to prevent a subnationalgroup from fabricating a nucleardevice from diverted plutonium.The goal is simple: select aphysical or chemical storageform that cannot be directlyused in fabricating a fissionweapon and that cannot beprocessed into a usable form.Unfortunately, there is no tech-nical solution that achieves thisresult.

ES&H issues encompass a spec-trum of issues that extend wellbeyond identification of hazardsand waste streams. A credibleevaluation must examine pos-sible methods of hazardremediation and weigh the ad-vantages of their implementa-tion against the disadvantages.An important means for mini-mizing ES&H risk is the avoid-ance of all unnecessary processoperations.



Credible interim storage optionsmust be technically feasible andcapable of near-term implemen-tation. Even if an option is tech-nically feasible, it may beimpractical due to excessivecosts or limited availability ofkey resources or material. Evalu-ation of a storage option is hin-dered by lack of knowledgeabout material properties, haz-ards, or process operations.Previous experience with agiven option is particularlyvaluable and is given appropri-ate consideration in this assess-ment.

Storage options that change thechemical form of the plutoniumhave especially far-reaching im-pacts on the total cost. In all like-lihood, the interim storage formwill not be identical to that cho-sen for final disposition. Addi-tional processing may ultimatelybe necessary. Interim conversionfrom one form to another ampli-fies this problem, particularly ifthe form is significantly diluted.The advantage of avoidingchemical processing is clearlyreflected in the cost because aunique facility is usually re-quired for each process. Suchfacilities are necessary becausemost chemical processes are cy-clic instead of reversible. For ex-ample, the technologies andprocess equipment used for con-verting metal to oxide are differ-ent from those for reducingoxide to metal. The processingfacility must first be configuredfor preparing the desired storage


A final consideration involvescontingency planning and pres-ervation of flexibility for finaldisposition. Certain physicaland chemical forms are moresuitable for a rapid, low-cost re-sponse than others. Maximumflexibility is achieved by storinga form that requires minimumprocessing for interim storage,but this ability to respond standsin conflict with a desire for in-creased proliferation resistance.Existence of the necessary tech-nology base for processing andverification within Complex 21are important factors.

The options examined for stor-age of excess plutonium fall intothree categories:

1) storage as intact pits,2) storage as altered pits, and3) storage as extracted

material.An assessment of each categoryis presented below.

Intact PitsStorage of plutonium as intactpits is an approach that employsthe intrinsic characteristics thatmade the pit suitable for ex-tended storage in a weapon con-figuration. These featuresinclude a critically safe pluto-nium geometry, a knownnuclear material inventory, anda durable, hermetically sealedcontainment vessel with certi-fied closures. A plutoniumfacility is not needed forimplementation of this optionbut is required for surveillanceand disassembly of pits that areidentified as unsuitable forcontinued storage.

Storage as intact pits has distinctadvantages for all technical is-sues. Hazards and ES&H risksare minimal. The pit is a well-certified containment vessel, andfeasibility is certain and sup-ported by an extensive and ex-cellent history of storage in theweapons stockpile. The desir-ability of intact pit storage is re-flected in the comments of along-time production workerfrom the Rocky Flat’s Plant:“We don’t build triggers fornuclear weapons. We build thesafest, most certified, best char-acterized storage containers forplutonium known to man.”This option preserves maximumflexibility and is the most cost-effective option due to limitedfacility requirements and theabsence of radioactive wastegeneration.

In terms of proliferation resis-tance, pit storage is often consid-ered highly undesirable becauseof the possibility for diversionand use in a nuclear device.However, the potential for utiliz-ing diverted plutonium fromother storage configurations iscomparable to that for an intactpit. The availability of intactpits for reconstruction of a largearsenal by a major nuclear statemay not be of concern becausethe production rate of weaponsystems is influenced by opera-tions other than pit fabrication.Indeed, experts in nonprolifera-tion and arms control point tothe fabrication of deliverysystems as the rate-limitingstep in the reconstruction of asuperpower’s nuclear arsenal.

Fabrication of the warhead haslittle if any impact on the timerequired to reconstitute a largestockpile. In fact, this very ob-servation is reflected in the nu-merous arms control treatiessigned by the United Statesand the former Soviet Union.START I, START II, and the INFtreaties all address the disposi-tion of delivery vehicles buthave no provisions or protocolsfor destruction of actual nuclearwarheads.

Storage of plutonium as intactpits is preferred by the scientificcommunity over all otheroptions under consideration.Nonetheless, the political costand perceptions associated withstorage of intact pits may over-ride the technical issues and pre-vent its ultimate implementation.

Altered PitsPit alteration is an approach thatfocuses on retaining advanta-geous pit features such asnuclear material containmentwhile destroying the suitabilityfor direct reuse. Two levels ofalteration are proposed:

1) modification of nucleargeometries by mechanical

or physical methods, and 2) modification of nuclear

geometries and material properties by chemical methods.

Whereas mechanical and physi-cal methods include deformationand insertion of inert materialsto prevent direct reuse in weap-ons fabrication, chemical meth-ods are designed to hinder ordeny reuse by also forming plu-tonium compounds that are diffi-cult to reconstitute to a usableform. Suggested chemical meth-ods include in situ formation of aproduct such as hydride, oxide,nitride, or alloy.



83

A detailed discussion of specificmethods is beyond the scope ofthis paper.

Alteration methods may be ca-pable of denying the rapid re-construction of an arsenal undervery specific scenarios, but theutility of alteration as a meansof preventing proliferation bynonnuclear states and terroristgroups is doubtful. To varyingdegrees, mechanical, physical,and chemical alteration methodsonly extend the length of timebetween diversion and reconsti-tution of material into a usableform. Estimates of the time re-quired for such reconstitutionhave been made, but in no casedo they provide a significantbarrier to extraction and purifi-cation of nuclear material. Thetechnologies required to reformmetal and to reprocess sug-gested chemical forms intometal are well known and canbe implemented at low cost, es-pecially if ES&H risks are disre-garded.

Although pit alteration methodsmay provide a small measure ofproliferation resistance (essen-tially in the political messagethey might send), this gain mustbe weighed against the disad-vantages. As noted previously,an important advantage of pitalteration is retention of the pitas a storage vessel; however,the certified integrity of thecontainer is destroyed by allmethods requiring access tothe pit interior. The potentialfor release of nuclear materialsduring those operations willundoubtedly require use of aplutonium-handling facilityand implementation of a certifi-cation procedure for reclosure.

Verification of pit disablementcould be accomplished by anumber of nonintrusive meth-ods that would not compromiseclassified information.

Risk and cost are significantlyincreased by use of chemicalmethods that involve injectionof reactive solids, liquids, orgases into the pit. Useful reac-tions must be complete in a rea-sonable time period to maintainprocess throughput. Difficultiesarise because of thermodynamicand kinetic factors. In somecases, the reactions are highlyexothermic (for example,∆H°f of PuO2 = -252 kcal/mol),and the likelihood of excessivetemperatures would requirecontrol of processing rate andheat flow to prevent thermallyinduced failure of pits as con-tainment vessels. Other reac-tions proceed at prohibitivelyslow rates, and excessively hightemperatures would be requiredto promote those processes.None of the proposed chemicalreactions has an appropriate rateat room temperature, and noneof the methods is sufficiently de-veloped to permit a thoroughassessment of hazards, assess-ment of stability during ex-tended storage, identification ofthe difficulties in reprocessingfor final disposition, or estima-tion of waste generation levels.

Extracted MaterialExtraction of plutonium de-stroys the pit configuration andpermits a broad range of candi-date forms to be considered forstorage. The absence of classi-fied shapes facilitates transpar-ency in storage, but verificationof material transfer during ex-traction remains a concern.

The thermodynamic and kineticcomplications cited for in situalteration are eliminated by flex-ibility in the design of processequipment. As with chemicalalteration, the desirability ofstoring selected material formsresides in the potential for pre-venting reuse and/or reprocess-ing of plutonium. Placement ofplutonium in a dilute form suchas a fused silica matrix or as amixture containing high-levelradioactive waste does not pre-vent its recovery by a motivatedgroup but would significantlyhinder further United Statesprocessing because of stringentES&H concerns. A decision tostore plutonium in such formsis essentially equivalent to se-lecting an option for final dispo-sition and must be carefullyconsidered as an option for in-terim storage.

A qualitative assessment ofissues for candidate storageforms of extracted plutoniumis presented in Table I. In addi-tion to the plutonium com-pounds mentioned in thediscussion of chemical alter-ation, the candidate storageforms include plutonium metal,plutonium dicarbide, and amixed plutonium-uranium ox-ide. Results of the evaluationare particularly significant;none of the candidate formsprevent proliferation, and differ-ences in time periods requiredfor reprocessing are small.




Candidate Proliferation Potential Remediation Anticipated Ease of SNM Storage Flexibility Forms Resistant Hazards Available Total Cost Accountability Experience

metal no yes yes low high extensive highoxide no yes yes moderate good high reducedmixed oxide no yes yes high reduced modest lowdicarbide no yes (?) no (?) high very low limited reducednitrite no yes (?) no (?) high low limited reducedhydride no yes* no low low limited highalloys no yes (?) no (?) high very low none low

(?) Insufficient data is available. *Candidate form is especially hazardous because of pyrophoricity.

from literature sources showsthat massive (greater than 0.2-mm-thick) plutonium becomespyrophoric only at temperaturesnear 500 °C.2 Plutonium finesand foils ignite when heated to150–200 °C and are not suitablefor storage. Oxidation of pluto-nium is prevented by storagein sealed metal containers withcertified closures.

The remaining potential hazardsidentified for storage of pluto-nium metal do not present seri-ous ES&H risks. Metal is not adispersible form of plutoniumand must be oxidized beforeentrainment is possible. The oxi-dation rate at room temperatureis such that dispersible particlesare released from clean metalonly after several days in moistair at room temperature.3 Above500 °C, air oxidation of a typical4-kg metal casting by the con-stant-rate process is completeafter several hours, but less than0.1 mass percent of the oxideproduct exists as particles withsizes in the dispersible range be-low 10-µm geometric diameter.

Table I. Qualitative Assessment of Issues for Candidate Storage Forms of Extracted Plutonium



Since all candidate forms havepotential ES&H hazards, the se-lection of suitable storage formshinges on the remaining issuesin Table I. The deciding factorsare availability of methods forhazard remediation, total cost,and storage experience. Theassessment of cost includesthe loss of flexibility and theeconomic impact of first process-ing for storage and then repro-cessing for final disposition.Cost is driven by the need todevelop both processing andreprocessing technologies, toinstall and replace process facili-ties, and to dispose of associatedradioactive waste. The low as-sessment of flexibility for mixedoxide and alloys arises becausereprocessing involves dissolu-tion in aqueous media and sub-sequent isolation of plutoniumby ion exchange or solventextraction. Other candidateforms are directly recoveredby pyrochemical methods.

Results in Table I suggest thattwo storage forms, metal and ox-ide, are the favored candidatesfor interim storage. The remain-ing discussion of extracted pluto-nium focuses on these materials.

Plutonium metal is an accepta-ble form for interim storage.Although metal is often viewedas a highly undesirable storageform because of proliferationconcerns, the technical assess-ment indicates that its prolifera-tion resistance is comparable tothat of oxide. Potential hazardsidentified for metal storageinclude chemical reactivity,dispersal of radioactive contami-nation, and pressurization of thestorage container. The primaryhazard is the chemical reactivityof metal in air and the potentialfor pyrophoric ignition.

The oxidation rate of plutoniummetal is modest and controlledby a protective oxide layer attemperatures up to 450 °C. Theoxidation rate in air is constant(0.2 g PuO2/cm2/min) at tem-peratures above 500 °C.1 Arecent analysis of ignition data

85

rate of helium formation fromdecay of plutonium-239 is welldefined.8 This decay product,which is retained as bubbles inmassive plutonium metal andreleased from oxide powder, in-creases the internal pressure ofan oxide container by about 1atm over a hundred-year period.An additional pressurizationhazard encountered duringshort-term storage and transpor-tation arises from desorption ofmolecular species at elevatedtemperatures created by self-heating or fire.3 Use of ventedstorage containers provides amethod for releasing pressureduring storage, but radiolysis ofair produces high concentrationsof gaseous nitrogen oxides thatare highly corrosive in the pres-ence of moisture.

Assessment of hazards associ-ated with plutonium storageindicates that the greatest envi-ronmental risk arises from dis-persal of process oxidefollowing a container ruptureor a spill.5 This conclusion isbased on particle size datashowing that 100 mass percentof a process oxide is in the dis-persible range. The dispersalrisk is a thousand times greaterfor process oxide than for theworst-case incident with metal.Possible methods of remediationinclude sintering of oxide pow-ders to increase the effectiveparticle size and hot-pressing ofoxide into briquettes; however,the cost and waste consequencesof these operations must be care-fully evaluated.

Hydrogen and other gasesformed by alpha radiolysis ofplastics and organic materialsin the storage environment arenot a hazardous pressurizationsource because they are getteredby reaction with plutonium.The products of these reactionsfrequently include plutoniumhydride and sesquioxide thatpresent pyrophoricity hazardsupon exposure to air. Ruptureof a storage container is possibleif it is not hermetically sealedand pressure is exerted by ex-pansion of the oxide product.

Assessments of the remainingissues for metal storage arefavorable. Experience gainedfrom storage of thousands ofpits defines the conditions re-quired for extended storage ofmetal and demonstrates thatsuch storage is remarkably freeof problems. SNM accountabil-ity is facile, maximum flexibilityis preserved, and minimal wasteis generated by storage of metal.The cost of a processing facilityis minimal because metal stor-age is compatible with thepyrochemical technologies ofthe Complex 21 baseline. If anyextracted form (metal, oxide, orother) is selected for storage, thethroughput capacity of the metalextraction facility must be ap-propriate to process the inven-tory of retired pits in a timelymanner that cannot exceed theprojected lifetime of the facility.

Plutonium dioxide is also anacceptable storage form. Theproliferation risk of oxide is con-sidered equivalent to that of plu-tonium because the metal is


readily reconstituted by meth-ods such as direct oxide reduc-tion (DOR) using commerciallyavailable chemicals and equip-ment. Plutonium dioxide is gen-erally considered to be a stableand inert material, but concernsarise because of the high surfacearea and tenacious adsorption ofmolecular species including wa-ter, organic compounds, and at-mospheric constituents. Processoxides formed by pyrolysis ofprecipitates such as oxalate, ni-trate, or peroxide have specificsurface areas from 20 to 60 m2

per gram4 and chemisorb sev-eral mass percent of water.5 Theremoval of adsorbates includingwater is accomplished by heat-ing the oxide at 1000 °C.6

Potential pressurization hazardsduring oxide storage arise fromgases generated by radiolyticand chemical reactions. Alpha-induced decomposition of ad-sorbates and organic materialsforms gaseous products includ-ing H2, CO, and HCl. A singlemonolayer of water, which canbe readily readsorbed on an ox-ide surface if adequate precau-tions are not taken after firing,is capable of generating 5 to 10atm of pressure in a typicalstorage container over time.3Although the dioxide is gener-ally considered to be thermody-namically stable, a recent studyshows that the dioxide reactswith water to form H2 andPuO2.2, a compound containingPu(VI).7 Although the kineticsof these radiolytic and chemicalprocesses are unknown and can-not be accurately predicted, the





Remaining issues related to ox-ide storage include nuclear ma-terials accountability, storageexperience, and flexibility. Tovarying degrees each of theseissues impacts the others andthe cost. Variations from 1 to 10mass percent are observed in theplutonium content of processoxides because of differences inthe amount of unpyrolyzed an-ions remaining after firing.4Although such complicationswith material accountability arewell known, storage experiencefor oxide is limited and proce-dures for outgassing and certify-ing oxides prior to storage arepoorly defined. If oxide isadopted as the only storageform for extracted plutonium,the processing facility must beappropriately sized and includean additional capability for effi-ciently converting extractedmetal to oxide. The aqueous re-covery technologies in the Com-plex 21 baseline produce oxidebut are designed for treatingresidues and pyrochemical salts,not for oxidizing large quantitiesof metal. If only oxide is stored,the potential impact on cost andwaste will be large.

ConclusionsThe selection of an option forinterim plutonium storage onthe basis of proliferation resis-tance cannot be justified on atechnical basis. Alteration mea-sures are considered effectivein forcing a nuclear state torefabricate pits before recon-structing a large arsenal, but thatdelay may not limit the rate atwhich weapons are produced.

The value of chemical alterationof pits and reaction of extractedplutonium to form proliferation-resistant material forms is ques-tionable because reconstitutionof candidate materials to usableforms is facile. Furthermore,most candidate forms couldprobably be used for construct-ing a nuclear device without be-ing reconstituted. Adequatesafeguards and security providethe only effective method forpreventing diversion and reuseof nuclear materials.

The formulation of a compre-hensive strategy is essential forachieving the goal of plutoniumprocessing with minimal waste.Storage of excess plutoniummust be included in that strat-egy. Development and imple-mentation of advancedtechnologies that reduce thequantities of radioactive wasteproduced and those that providecapability for decontaminatingwaste are essential componentsof the strategy. The maximumimpact on waste generation maybe realized by applying a simpleintuitive approach: avoid all un-necessary processing. Conse-quently, plutonium that is to beplaced in interim storage shouldremain in its existing form; re-processing and interconversionof chemical forms should beavoided unless a potentiallyhazardous situation exists.

A comprehensive strategy mustalso consider forms of pluto-nium that are not part of thepresent assessment. Such mate-rials include incinerator ash,pyrochemical salts, anode heels,etc., that are presently in storageat several DOE sites. The situa-tion is extremely complex andapplication of general guidelinesmay not be appropriate. Furtherevaluation of these areas isclearly needed.

The present technical assess-ment provides a basis for gen-eral recommendations oninterim storage of plutonium.Both metal and oxide are accept-able storage forms. Storage ofmetal as intact pits is preferred.In the absence of a suitable plu-tonium handling facility, intactstorage is the only near-term op-tion. If interim storage of intactpits is deemed unacceptable,storage as extracted metal is pre-ferred. Storage as altered pits isthe least desirable option. If al-teration of pits is necessary, onlymechanical or physical methodsthat preserve the integrity ofcontainment are acceptable.Metal and process oxide shouldbe stored in their existing formsafter preparation and certifica-tion for storage. All storagevessels should be hermeticallysealed and should exclude or-ganic and other covalentlybound materials. Regardless ofthe storage option(s) employed,adequate and continuing sur-veillance will be necessary toidentify potential problems.

87

7. J. L. Stakebake, D. T. Larson, and J. M. Haschke, “Characterization ofthe Plutonium-Water Reaction PartII: Formation of a Binary OxideContaining Pu(VI),” J. of Alloys andCompounds 202, 251 (1993).

8. J. C. Martz, “Analysis of Pluto-nium Storage Pressure Rise,” inNMT-5:92-328, July 1992, LosAlamos National Laboratorymemo.

Cited References1. J. M. Haschke, “Evaluation ofSource-Term Data for PlutoniumAerosolization,” Los AlamosNational Laboratory report LA-12315-MS (July 1992).

2. J. C. Martz, J. M. Haschke, andJ. L. Stakebake, “A Mechanism forPlutonium Pyrophoricity,” LosAlamos National Laboratory reportLA-UR-93-2655, submitted forpublication, July 1993.

3. J. M. Haschke and J. C. Martz,“Metal-Oxide Chemistry andStorage,” Los Alamos NationalLaboratory report LA-CP-93-159(May 1993).

4. J. D. Moseley and R. O. Wing,“Properties of Plutonium Dioxide,”Dow Chemical Company reportRFP-503 (August 1965).

5. D. Y. Chung, et al., “Assessmentof Plutonium Storage Safety Issuesat Department of Energy Facilities,”Department of Energy draft report,Office of Defense Programs,Germantown, MD, September 1993.

6. J. L. Stakebake and L. M. Stew-ard, J. Colloid Interface Sci. 42, 581(1973) .

Several technology areas needfurther definition and develop-ment. Continuation of efforts toupgrade process technologies forefficiency and waste minimiza-tion are essential. Additionalwork on the kinetic and equilib-rium behavior of oxide adsorp-tion is needed with emphasison the conditions required forpreparing, certifying, and pack-aging material for storage.Studies to define the chemicaland radiolytic reactions betweenwater and oxide are needed topredict long-term storage behav-ior. The merits of removing am-ericium from extracted materialprior to storage also requiresevaluation. The storage ofplutonium-containing residuesneeds to be addressed, and theconsistency of storage decisionsand Complex 21 design mustbe monitored.

In addition to providing a tech-nical basis for selection of stor-age options, the present studyhelps to place waste issues andthe role of technology develop-ment in perspective. Althoughdevelopment of new technolo-gies for minimization and pro-cessing of radioactive waste isnecessary, development of tech-nical input for policy decisions isalso essential to ensure that theobjectives are consistent andcomplementary.✦




Waste Management at Technical Area 55: A Historical Perspectiveand Status ReportBill J. McKerley,James J. Balkey,and Ronald E. WienekeNuclear Materials Processing:Nitrate Systems

IntroductionIn August 1992, the transuranic (TRU) waste management opera-tions for the Nuclear Materials Technology (NMT) Division weretransferred into the Nitrate Systems Group (NMT-2), one of nineoperating units within the Division. In February 1993, all wastemanagement activities at Technical Area 55 (TA-55) were consoli-dated. The areas of immediate concern were backlogged legacywaste items (low-level, TRU, and hazardous) stored in the base-ment of the Los Alamos Plutonium Facility in TA-55, cementedTRU waste drums stored at an off-site location, experimental radio-actively contaminated vessels stored on the Plutonium Facilitygrounds, and oversized waste items stored in the Plutonium Facil-ity basement. Communication was initiated with the Laboratory’sWaste Management Group (CST-7) to discuss options for resolvingthese issues. This report describes the technical, regulatory, andadministrative actions and accomplishments that allowedNMT Division to effectively address these important issues.Activities in the pursuance of waste minimization goals are alsoreviewed.

BackgroundWaste management at the Plutonium Facility was one function ofthe old Nuclear Materials Management Group (NMT-7). The groupwas responsible for such diverse activities as shipping and receiv-ing, nuclear material storage, roasting and blending of reactor fuels,and the development of a site-wide nuclear materials model (insupport of Complex 21 planning) to forecast the impact of newtechnologies on scrap inventories and waste generation. Withlimited personnel resources and diverse operations, group staffcould not keep up with the steadily increasing requirements anddemands of waste management regulations. Likewise, personnelresources were not available to assign additional technicians towaste management activities as needed to address the resultingoperational demands and problems. Thus, waste managementpersonnel were faced with a variety of seemingly insoluble prob-lems without the resources to explore and implement solutions.

Waste Management IssuesA variety of difficulties in several operations had developed in thewaste management program at the Plutonium Facility. These prob-lems, outlined below, threatened waste certification and shipmentof TRU, mixed, and low-level waste to the Laboratory’s interimstorage facility and low-level waste disposal site, TA-54.

“The areas of immediateconcern were backloggedlegacy waste items (low-level, TRU, and hazardous)stored in the basement of theLos Alamos PlutoniumFacility at TA-55, cementedTRU waste drums stored atan off-site location, experi-mental radioactively con-taminated vessels stored inthe Plutonium Facilityyard, and oversized wasteitems stored in the Pluto-nium Facility basement.”

Facility Operations

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Facility Operations

Waste Management at Technical Area 55: A Historical Perspective and Status Report

Lack of Suitable NondestructiveAssay Capabilities for 235UWaste. Another complex issueinvolved the nondestructiveassay of waste packagescontaining uranium-235 as aco-contaminant with plutoniumfrom mixed oxide fuel process-ing operations. At the time ofthis waste generation, no suit-able nondestructive assay capa-bility existed at the PlutoniumFacility. But the SafeguardsAssay Group (N-1), who is thevanguard of nuclear materialdetection technology develop-ment at the Laboratory, assayedabout two dozen of these drumsfor uranium-235 content. Thisoperation was outside of thedomain of the Laboratory TRUwaste program and, conse-quently, outside of the integralquality assurance program.This technicality had to beresolved before the uraniumSpecial Nuclear Material (SNM)assays could be judged accept-able and waste could be certifiedfor storage at TA-54.

Nonconforming 239Pu WasteDrums. An inventory of twodozen nonconforming pluto-nium-239 waste drums (due toeither SNM or weight compari-son discrepancies) had accumu-lated in basement storage areasof the Plutonium Facility. Theparticular discrepancies couldnot be investigated and resolveddue to limited staffing and thepressures of keeping up withthe steady stream of waste gen-erated from plutonium process-ing operations.

Accumulation of Oversized TRUWaste. An accumulation ofoversized waste items (biggerthan would fit in 55-gallondrums), requiring either sizereduction or packaging in stan-dard waste boxes, filled storageareas in the basement of the Plu-tonium Facility. Shortage of per-sonnel available to support theTRU oversized waste operationalso contributed to this wasteproblem.

Accumulation of ContaminatedExperimental Vessels. A collec-tion of radioactively contami-nated experimental vesselsdating back to the 1960s was be-ing laboriously decontaminatedby a team of NMT-2 techniciansand was accumulating in theyard of TA-55. Methods of pre-paring the vessels for transpor-tation and storage at TA-54 hadnot been established with theCST-7 Waste ManagementGroup. Additional requirementsthat had to be met to transfer thevessels out of the PlutoniumFacility needed to be negotiatedas well.

Accumulation of ContaminatedGlove Boxes. For four years,glove boxes from decontamina-tion and decommissioningoperations in the PlutoniumFacility had been crated in ply-wood crates and moved out intothe TA-55 yard for interim stor-age. As in the case of the experi-mental vessels, additionaldocumentation or preparationthat would be required to movethe glove boxes off-site had notbeen established.

Water Release from CementedTRU Waste. In June of 1989,free liquid was found on thetop of a drum of cemented TRUwaste in the basement of thePlutonium Facility. Analysisindicated that this liquidworked its way up through thecarbon filter from the inside ofthe drum. This clear violationof the Waste Acceptance Criteriafor the Waste Isolation PilotPlant (WIPP), criteria that barfree liquids from waste pack-ages, immediately brought intoquestion the integrity of the 350waste drums that had been im-mobilized using the gypsumcement process.

Prohibitive Packaging Limitsfor 238Pu Waste. Transportationsafety requires that materialsmoved by highway, in this caseTRU wastes transported in theTRUPACT II overpack (Depart-ment of Transportation Type Bshipping package), must notcontain flammable or explosivemixtures of gasses. A primarysource of flammable gas in TRUwaste is hydrogen from alpharadiolysis of hydrogenous mate-rials in the waste matrix. Pluto-nium-238 has the highest spe-cific activity (which is directlyproportional to the gas genera-tion rate) of the plutonium iso-topes handled at TA-55 and,consequently, has the lowestallowable concentration limitsin WIPP waste packages. Pack-aged TRU wastes, especiallythose in multilayer packagingconfigurations, did not meet theWaste Acceptance Criteria forWIPP and, therefore, were notcertifiable or suitable for storageat TA-54.



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The destination of the gloveboxes would be the TA-50 SizeReduction Facility where theywould be cut into pieces andconsequently repackaged inacceptable waste containers(55-gallon drums or standardwaste boxes). Water had accu-mulated in some boxes fromprolonged exposure to the ele-ments and would have to beanalyzed and removed fromthe packages before shipment.

Accumulation of Low-LevelOversized Waste. Low-leveloversized waste, packaged in4-ft x 4-ft x 8-ft plywood boxes,had likewise accumulated in theTA-55 yard, totaling thirty-sixboxes. These waste packagesadded to the congestion of mate-rials in the yard and occupiedvaluable space.

Thus, waste items were stallingin the “pipeline” that processed,packaged, and funneled wasteitems to the waste managementstorage and disposal site atTA-54. As a result of these diffi-culties, waste operations atTA-55 were coming to a haltand were threatening the re-search programs that generatedthe waste. Resources had to bereallocated immediately to alle-viate this situation.

New Regulatory SettingIn the spring of 1992 both theDepartment of Energy (DOE)and the Environmental Protec-tion Agency (EPA) realized thatan agreement must be reachedconcerning Resource Conserva-tion and Recovery Act (RCRA)compliance (which dealt with

hazardous waste issues) at DOEWeapons Complex sites; there-fore, a commitment to draft aFederal Facility ComplianceAgreement (FFCA) was made.The DOE also made an agree-ment with the EPA to ceasegenerating hazardous and, con-sequently, mixed wastes from alldefense operations. This deci-sion effectively shut down mostradiochemical processes at thePlutonium Facility as of May 8,1992. The moratorium on mixedwaste generation presented botha challenge to and an opportu-nity for the TA-55 mission. At atime when waste managementneeded manpower to resolvegrowing waste problems, pro-cessing operations were cur-tailed, resulting in idledpersonnel that, potentially,could be reassigned to addressthis need.

Group ReorganizationDivision resources were imme-diately put to work on wastemanagement problems at TA-55.The TRU liquid and solid wasteoperations were transferred tothe NMT-2 Nitrate SystemsGroup and placed under the Ni-trate Processing Team Leaders.This decision was based uponthe fact that NMT-2 producedmost of the waste being pro-cessed, and nitrate operationspersonnel had intimate knowl-edge of the waste produced.NMT-2 group management alsohad a no-nonsense reputationfor defining and solving prob-lems; furthermore, many of thepersonnel idled by the morato-rium on mixed waste generationwere NMT-2 employees.

Responsibility for the PlutoniumFacility basement storage andlow-level waste activities andthe TA-55 yard storage areaswas transferred to the NMT-8Facilities Management Group,since these functions matchedNMT-8’s mission better. Andthe shipping and receiving op-erations were transferred to theNMT-4 Nuclear Materials Mea-surements and AccountabilityGroup.

Not all aspects of the reorganiza-tion were positive. The dissolu-tion of the NMT-7 NuclearMaterials Management Groupresulted in time-consuming at-tempts to coordinate interrelatedFacility waste management op-erations that were now locatedin separate Laboratory hostgroups. The reorganization alsoresulted in the loss of experi-enced waste management per-sonnel; out of thirty personneloriginally in the NMT-7 Group,approximately half chose totransfer to other types of posi-tions rather than move with thewaste management functionsassigned to other NMT groups.New technicians and staff hiredto replace them required site-and operation-specific trainingbefore assuming their duties.

Waste Management SolutionsThe concentration of NMT-2 re-sources on waste managementproblems at TA-55 over the pastyear and a half has yielded dra-matic results. Waste manage-ment operations are staffing upto realistic levels to handle theworkload, and personnel are be-ing certified in their operations.

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Group for a waiting period of84 weeks before shipment toTA-54, the approximate time forthe cement curing process basedupon earlier research by NMTpersonnel. Waste managementpersonnel would then visuallyconfirm the absence of free liq-uids before shipment to TA-54.Any packages in which moisturewas observed or that could notbe inspected (such as drumscontaining one-gallon cans)would require overpacking.The drums stored in the CMRbuilding were visually exam-ined where possible and thenoverpacked in polyethylene-lined 83-gallon drums, and theannulus was filled with absor-bent materials. CMR repackag-ing operations were conductedin September and October 1993;shipment of all 218 of the CMRdrums was completed the firstweek of October. To date, 107drums in the Plutonium Facilitybasement storage areas havebeen held the requisite time, vi-sually examined, and shipped toTA-54 for storage; the remainingdrums should be shipped outshortly after the beginning of1994.

Research on the gypsum cementprocess and on the alternativereturn to portland cement is be-ing conducted. Research intothe portland cement process re-quires addressing the control ofprocess-related ammonium ni-trate emissions. Experimentalwork is also being conducted ontoxic metal retention, since theevaporator bottoms (the radioac-tive residues resulting fromevaporation) contain RCRA-regulated concentrations ofchromium, lead, and nickel.

Dialog has been reestablishedwith the Laboratory waste man-agement and environmentalcompliance organizations, pro-viding a means of resolvingproblems in a mutually agree-able manner. And legacy wasteissues are being resolved: oldwastes are being remediated andsent to TA-54 for storage, andprocedures are being modifiedto avoid similar future prob-lems. Solutions to the wastemanagement problems outlinedearlier in this paper are givenbelow.

Repackaging of Cemented TRUWaste Drums. Drums of ce-mented TRU waste suspectedof containing free liquid werereturned to TA-55 (point of gen-eration) from interim storage atTA-54. Because there was notenough room for the returningdrums in the Plutonium Facilitybasement, additional space fordrum storage was arranged atthe Chemical and MaterialResearch (CMR) building. Thefirst challenge was to determinethe extent of the problem; thenext challenge was to find thecause of the water release fromthe cement monoliths. Threedrums were found to containsignificant quantities of freewater attributable to variancesin the cement mixing process.Small quantities of water vaporare normally released duringthe cement curing process bythe heat of hydration and some-times condense on the inside ofthe plastic liner. Samples of theliquid were found to containvery little radioactivity. Theplutonium is effectively immobi-lized in the cement, even thoughfree liquid may appear. NMTpersonnel negotiated with theCST-7 Waste Management

Safe Storage Limits for 238PuWaste Drums. The NMT-9 HeatSource Technology Group hasbeen conducting a research effortto determine the safe quantity ofplutonium-238 allowed in wastepackages. They have been re-searching existing literature,conducting experiments undercontrolled conditions, and doingheadspace gas sampling (withthe cooperation of NMT-2 wastemanagement personnel) onlegacy waste packages in thebasement of the Plutonium Facil-ity. Work that involves CST-1Analytical Chemistry and CST-7Waste Management and Bench-mark will establish a reasonablestorage limit that will ensure safestorage of plutonium-238 con-taining waste but that will exceedconservative TRUPACT-II Con-tent Code (TRUCON) limits forshipment to WIPP. Storage re-quirements will be based uponwaste matrix and containmentlayers. Legacy waste drums arebeing opened to confirm packag-ing configuration, and the first 10drums of plutonium-238 wasteare being prepared for shipmentto TA-54 for interim storage.

Nondestructive Assay Capabili-ties for 235U at TA-55. The WIPPwaste program allows evaluationof certifiable wastes by instru-ments that can be proven to beequal in integrity to those usedfor waste assay within the ap-proved TRU Waste CertificationProgram, including its integralquality assurance program.The N-1 nondestructive assayinstrumentation and methodsused in the uranium assay ofthe mixed plutonium/uraniumdrums were rated as equivalentto those used in the TRUWaste Certification Program.


limited analysis capacity andlow priority. Some of thesamples took a year to be ana-lyzed. Nondestructive assay ofthe vessels was challenging dueto the thick steel walls and con-sequent attenuation of the radia-tion used to detect and quantifythe radioisotopes in the waste.NMT-4 ran calibration tests anddetermined uncertainties basedupon a series of experimentsconducted on an uncontami-nated vessel with a knownsource placed inside. This infor-mation was used to devise thecounting scheme for the vessels,and the clean vessels were thenassayed. Calculations weremade based upon the weightsof the vessels, the results of theRCRA constituents’ analysis,and the nondestructive assay.Then the vessels were classifiedfor disposal. In September,twelve 3-ft-diameter vesselswere shipped to TA-54 for in-terim storage until the TA-50Size Reduction Facility is able toprocess them. They will eventu-ally be placed inside standardwaste boxes, which are accept-able primary containers forTRUPACT II shipment to WIPP.

Removal of ContaminatedGlove Boxes and Oversized,Low-Level Waste Boxes. Theproblem with the glove boxesand the 4-ft x 4-ft x 8-ft low-levelwaste boxes centered upon as-sembly of the proper documen-tation necessary for review andacceptance by the CST-7 WasteManagement personnel. Afterconsultation with CST-7, paper-work packages were preparedby NMT-2 waste managementpersonnel, and shipments werescheduled. Some of the glove

These assays were then used tocomplete the paperwork pack-ages, which were circulated forreview and approval. In addi-tion, the NMT-4 NuclearMaterials Measurements andAccountability Group com-pleted programming and certifi-cation of one of the SegmentedGamma Scanning instruments,capable of assaying 55-gallonwaste drums for uranium. Thisinstrument was used to assaydrums with low atomic weightwaste matrices. Approximatelyhalf of the 40 mixed uranium/plutonium waste drums wereassayed, the paperwork pack-ages were completed, and thewaste drums were sent to TA-54for storage.

Inspection/Repackaging of239Pu Waste Drums.Nonconfirming plutonium-239drums were organized into twomajor groups: (1) those forwhich the final gross drumweights did not match the sumof the individual item weightsand the package tare (emptycontainer weight), and (2) thosefor which the total drum SNMassay did not agree with thesum of the individual item SNMassays. Paper work was first ex-amined for errors; if none couldbe found, then the drums werecarefully opened and the itemsreweighed or reassayed as ap-propriate until the discrepantitem was found. The packagepaperwork was then correctedand routed for reviews and ap-provals, allowing shipment toTA-54 for interim storage.


Decontamination/Repackagingof Oversized TRU Waste. Workis progressing on the disposal ofoversized TRU waste itemsstored in the basement of thePlutonium Facility. In early1993, a concerted effort wasmade to identify and categorizewaste items in this area. Ap-proximately a dozen items weredecontaminated, repackaged,and disposed of as low-levelwaste. Seven items that weretoo large to fit in standard wasteboxes were crated and are in theprocess of being assayed forSNM content by NMT-4. Theywill be shipped to TA-54 forinterim storage pending theavailability of the TA-50 SizeReduction Facility for process-ing waste.

Analysis/Decontamination ofExperimental Vessels. The de-contamination work being con-ducted by NMT-2 personnel onthe experimental vessels hasmade steady progress, and atpresent, twenty vessels havebeen decontaminated. Twomajor determinations that hadto be made prior to disposalwere these:

(1) whether the vessel waste residues were TRU or low-level waste and(2) whether the residues included mixed waste.

To make these decisions, NMT-2personnel assumed that the con-tamination on the inside of thevessels was uniform and that thewaste residues removed wererepresentative of the contamina-tion. Samples of these residueswere sent to CST-1 for analysis,but analysis was delayed due to

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boxes that had been in storageand exposed to the environ-ment for several years had waterinside the crates, which had tobe drained in a manner accept-able to waste managementand health physics personnel.The thirty-six 4-ft x 4-ft x 8-ftboxes were shipped to TA-54 inMarch 1993, freeing considerablespace in the yard for other pur-poses. The glove boxes, whichrequired more preparation thanthe low-level waste boxes, werenot shipped off-site until July1993.

Waste MinimizationWaste in any form is expensiveto package, document, treat, anddispose of; costs are escalatingas regulations become morestringent, treatments becomemore exotic, and landfill spacedisappears. Waste minimizationserves to reduce the expense ofwaste operations by avoiding orminimizing waste in the firstplace. It is also mandated byDOE Order 5820.2A, “Radioac-tive Waste Management,” andthe Resource Conservation andRecovery Act.

The philosophy of waste mini-mization at the Los Alamos Na-tional Laboratory is defined byAdministrative Procedure 10-8,“Waste Minimization.” To date,waste minimization efforts atTA-55 have been the responsibil-ity of the waste generators. Per-sonnel are encouraged in wastegenerator training to utilizewaste minimization as the mostdesirable method of waste man-agement in their operations.The generator has the responsi-bility to reduce waste not only inevery day operations but also indesigning future processes.

EpilogueDramatic progress has beenmade in the waste managementprogram at TA-55 over the pastyear. Reorganization of thewaste management team andsupport by NMT-2 operationspersonnel and management hasrestored operational viability.Communication, vital in dealingwith continually changing rulesand regulations, has been rees-tablished and enhanced. Prob-lematic legacy wastes have beenaddressed and remediatedwhere possible, and the rootcauses have been identified toprevent recurrence. The numberof technical staff and operationstechnicians is being increased tohandle the workload, and newpersonnel are being trained onboth regulations and practicalaspects of operations. Muchwork has been accomplishedover the past year, but evenmore waste issues are yet to beresolved. The waste manage-ment organization must learn toanticipate mandated changesand must plan for facility opera-tion changes in such a way as tominimize the impact and burdento the primary objective of thePlutonium Facility: to provideto the DOE Weapons Complexfirst-rate research and develop-ment in actinide processing.♦

In the past, the need for wasteminimization was addressed ineach Safe Operating Procedure(SOP) as a self-contained sec-tion, but no coordinated effortwas made to provide compre-hensive and consistent guidanceto all operations at TA-55.

Currently, a waste minimizationplan for TA-55 is being written.This plan will provide guidanceto all personnel (from officeworkers to plutonium process-ing technicians) whose opera-tions generate waste on-site.The redundancy of addressingwaste minimization in eachSOP will be eliminated, andwaste minimization informationfor all personnel and all opera-tions will be consolidated intoone source that can be revisedexpeditiously as warranted.

Performance indicators (PIs) arebeing investigated that wouldprovide a measure of wasteminimization at the PlutoniumFacility. Information on pastprogram performance has beenlargely qualitative because onlyinformation on waste leavingthat facility has been recorded.Incorporation of PIs at TA-55 tomeasure waste minimizationwill require the commitment ofmanpower and computer timeto track the requisite variablesthat comprise each PI.


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Future of Waste Management at the Los Alamos Plutonium Facility

“We must find innovativeways of complying withenvironmental regulationswhile maintaining flexibil-ity of operations and anemphasis on research.”

Ronald E. WienekeNuclear Materials Processing:Nitrate Systems

IntroductionWith the enactment of the Federal Facility Compliance Act of 1992(FFCA), the national laboratories within the Department of Energy(DOE) weapons complex are no longer immune to fines for violat-ing Environmental Protection Agency (EPA) hazardous waste regu-lations. The Federal Facility Compliance Agreement between thelaboratories and their respective states will establish a timetable toachieve full compliance with existing requirements. If the agreed-upon milestones are not met, facilities and managers will be subjectto civil and possibly criminal penalties for violations. Those whooversee operations that generate hazardous waste must movequickly to implement waste management systems, but restraintand good judgment will be required to avoid overburdeningoperations with excessive documentation.

The Plutonium Facility at Los Alamos National Laboratory strivesto meet existing environmental regulations and to anticipate futurerequirements. We must find innovative ways of complying with en-vironmental regulations while maintaining flexibility of operationsand an emphasis on research. This enormous task can only be ac-complished with a spirit of cooperation among those who generate,manage, and regulate waste. This requires cooperation among(1) waste generators and waste management personnel within thePlutonium Facility, (2) waste management and environmental com-pliance groups at Los Alamos, and (3) external regulatory agencies.A reasonable approach to the interpretation of regulations as theyapply to the complexities of radiochemical processing operations isrequired at all levels.

The Plutonium Facility must look ahead and anticipate changes inrequirements for the management of hazardous waste. Carefulplanning will be needed to give us time to act in harmony with ourmission, rather than being forced to react to new requirements inways that merely minimize adverse consequences.

Deadlines for Proper Storage and Treatment of Los AlamosMixed WasteThe Federal Facility Compliance Act (FFCA) waives the doctrine ofsovereign immunity for federal facilities, such as DOE contractors,with respect to enforcement of environmental regulations. Thiswaiver allows civil and possible criminal penalties to be levied bythe EPA. The FFCA in essence gives the EPA the teeth that it needsto enforce compliance with the Resource Conservation and Recov-ery Act (RCRA) for hazardous and mixed waste operations. Fur-thermore, the DOE has agreed to negotiate agreements with therespective state environmental agencies on how to implementRCRA at each facility.

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Los Alamos must certify compli-ance with the storage criteria fortransuranic (TRU) mixed wasteby the end of January 1995 andmust present a plan to the NewMexico Environmental Depart-ment (NMED) to establish treat-ment priorities for low-levelmixed waste by the end of May1995. The treatment and dis-posal of this waste is signifi-cantly complicated by thepresence of radioactivity, as hasalready been demonstrated bythe problems encountered withstored mixed waste at otherDOE facilities. The ultimatetreatment and disposal of mixedwaste, which must be accom-plished in a manner that pro-tects employees, the public, andthe environment, presents botha technological and financialchallenge to the entire DOEweapons complex.

Cost and Complexity of WasteDisposalEach kind of waste generated atthe Plutonium Facility has itsown set of complex require-ments that must be met to allowdisposal. Ever growing require-ments will continue to escalatethe costs related to characteriza-tion and disposal of waste. Evenlow-level radioactive waste, arelatively uncomplicated wastestream, will be more difficultand costly to dispose of in com-ing years as the Los Alamoslow-level waste acceptance crite-ria become more stringent andlandfill space is depleted. Allpersonnel who generate wastemust assure that no RCRA haz-ardous components have beeninadvertently included in theirlow-level radioactive waste andmust demonstrate control overtheir waste streams.

waste streams. This documenta-tion is most effective when theprocess is very well defined andthe composition of incomingprocess streams is well known.Documentation saves a signifi-cant amount of money by elimi-nating analytical tests forcompounds that have not clearlybeen associated with the pro-cess. On the other hand, it is par-ticularly damning when ahazardous compound or condi-tion is discovered that was notdisclosed in the process knowl-edge proclamation. Such a dis-covery throws a shadow ofdoubt over all wastes so charac-terized. For this reason, controlover process knowledge docu-mentation must be rigorouslyexercised.

The Plutonium Facility mustdocument its processes carefully.All radioactive wastes must beprecisely characterized to verifythat no hazardous constituentsother than radioactivity arepresent. DOE Order 5820.2A,“Radioactive Waste Manage-ment,” requires that waste dis-posal facilities know thecontents of the waste that theyare burying by establishingwaste acceptance criteria forthose sites that generate andship the waste to them for dis-posal. In addition, generatorsmust also know the hazardouscomponent of the waste so thatappropriate handling, storage,and treatment methods can beimplemented.

Mixed waste, waste that hasboth radioactive and hazardouscomponents, is a much largerproblem. DOE is severely lim-ited in its ability to processmixed waste to meet the EPAwaste treatment standards dueto a finite number of viablemixed waste treatment optionsand a shortage of accessibletreatment facilities. Despitethese problems, waste must betreated in accordance with theEPA standards before disposal.Low-level mixed waste is re-quired to be treated in accor-dance with 40 CFR Part-268,“Land Disposal Restrictions,”and disposed of in speciallydesigned hazardous waste land-fills. The landfills must incorpo-rate all specified design featuressuch as double liners, leachatecollection and treatment sys-tems, a leak detection system,and monitoring wells.

Disposal of TRU mixed wasteshould be easier, provided theWaste Isolation Pilot Plant(WIPP) is successful in obtaininga no-migration variance for thedisposal phase. WIPP currentlyhas a no-migration determina-tion from the EPA for the testphase only. The disposal ofmixed waste (both low-leveland TRU) is costly, and thosewho generate waste must avoidthe production of additionalmixed waste whenever possible.

Documentation of WasteThrough Process KnowledgeProcess knowledge is the docu-mented evidence provided bycognizant operating personnelof the presence or absence ofhazardous compounds in their


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Mixed waste must be carefullyanalyzed for both hazardousand radioactive constituents be-cause disposal facilities mustsatisfy two sets of disposal sitecriteria. One of these is the pre-scriptive criteria of RCRA,which covers the hazardouscomponent of the waste. Theother is the performance-basedcriteria of 40 CFR 191, “Environ-mental Radiation ProtectionStandards for Management andDisposal of Spent Nuclear Fuel,High-Level and Transuranic Ra-dioactive Wastes,” which coversthe radioactive components inthe waste. To meet both of thesecriteria for a waste from an un-known source, the generatorwould have to perform a fullspectrum RCRA analysis onmixed waste in accordance withEPA’s SW-846, “Test Methods forEvaluating Solid Waste,” as wellas perform a comprehensiveradioassay. Estimated cost ofthe RCRA analysis alone for amixed waste could reach tens ofthousands of dollars per sample.This estimate is not unreason-able considering the possiblenumber of hazardous compo-nents, employment of tech-niques to reduce radiationexposure to operators, and thequality assurance requirementsfor samples and sampling proto-cols (including chain of cus-tody). Process knowledgedocumentation is, therefore, anextremely important consider-ation from the aspect of costreduction.

Impact of Waste Managementon ResearchAs the requirements of wastemanagement at the PlutoniumFacility grow, the program mustbecome more efficient. The Plu-tonium Facility mission is toperform basic and applied re-search in actinide chemistry andmaterials research in support ofplutonium processing needswithin the DOE weapons com-plex. Research at the PlutoniumFacility benefits current opera-tions involving cleanup and re-covery of plutonium-bearingscrap, safe decommissioningof weapons components, andpreparations for long-term Spe-cial Nuclear Material (SNM)storage. Research at the Pluto-nium Facility also provides thebasis for the design effort of thenext generation of DOE pluto-nium manufacturing facilities(Complex 21).

Meanwhile, requirements forhandling, storing, and certifyingwaste are increasing theworkload of process operationspersonnel. More time is requiredto complete the necessary docu-mentation and maintain controlover waste collection and stor-age at the Plutonium Facility.Every dollar that is spent for thedocumentation, treatment, han-dling, packaging, or disposal ofwaste is a dollar that cannot beused for research. In this age ofshrinking research budgets andgrowing competition for assign-ments and funding betweensites in the DOE complex, mak-ing optimum use of every fund-ing dollar is essential. For thisreason, new programs are beingplanned and initiated that willincrease the effectiveness of thewaste management operationsat the Plutonium Facility.

Problems with ConflictingRegulationsThe complexity of the opera-tions at the Plutonium Facilitycan bring regulations from dif-ferent agencies into conflict witheach other. For example, whatone agency may consider radio-active waste, another may viewas valuable, recoverable scrapmaterial. A significant portionof the material stored at thePlutonium Facility contains eco-nomically recoverable quantitiesof plutonium, but it may alsofall under the RCRA definitionof waste. This material, which isknown as “scrap” with respectto plutonium recovery, is beingstored until processing canresume. This situation has beenaggravated by the changingmission of the DOE complex,which has brought the conceptof economic discard limits forplutonium into question. Theabsence of a demand for pluto-nium in weapons productionand the surplus from the decom-missioning of retired weaponshave resulted in plutonium be-ing considered by some as aliability.

When the requirements of theAtomic Energy Act (AEA) andthe RCRA are both applied tomixed wastes, the result is oftenconflicting requirements. Thestorage requirements for radio-active scrap material classifiedas mixed waste present particu-lar challenges: the AEA storagecriteria for SNM emphasize se-curity, personnel radiationsafety, and criticality safety.


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Several of these safety principlesfor handling fissile materials arein conflict with EPA guidanceand RCRA requirements forhazardous waste storage. Spe-cifically, the periodic inspectionof items and the labeling of ex-isting stored containers as re-quired by the EPA increasepersonnel exposure significantly,and the EPA requirement to con-fine water used in putting outfires in storage areas couldpresent a criticality hazard.

Another example of conflictingguidance is the TRUPACT-IIContent Code (TRUCON) limi-tations, Radioactive MaterialManagement Areas (RMMAs),and waste minimization require-ments. Radioactive wastes aredisposed of on a volumetric costbasis; therefore, it is beneficial tominimize waste volumes to de-crease disposal costs, includingtransportation and handling.TRUCON limits the quantity ofplutonium-238 to 0.049 gramsper drum by specifying the al-lowable activity in organic wastecontained within a 55-gallondrum with three levels of con-tainment. This limit, which isbased on what many consider tobe an overly conservative trans-portation safety analysis, mayonly amount to one contami-nated tissue in a 55-gallon drum.

The premise behind the RMMAconcept is that all waste comingfrom a radiation control area ispotentially contaminated andshould be disposed of as radio-active. This premise is in directconflict with waste minimiza-tion principles. Because neitherthe DOE nor EPA has issuedde minimis radiation standards,nuclear facilities have had to usethe strict interpretation of a “noradioactivity added” doctrinefor the segregation of wastestreams. This overly conserva-tive stance will result in thegeneration of more suspectradioactive waste than is neces-sary. Solutions that will allowreasonable operating parametersto be implemented without com-promising safety and environ-mental concerns must be found.

Improving Communication andInteraction at Los AlamosCommunication must be en-hanced and encouraged amongpersonnel from the variousLos Alamos organizations thatare engaged in waste manage-ment activities. Waste manage-ment personnel in the Pluto-nium Facility have been forcedto interact closely with Labora-tory waste management andenvironmental managementgroups because of the numberand complexity of regulationspertaining to radioactive, haz-ardous, and mixed wastes.This interaction has resulted inseveral benefits to the Labora-tory and the Plutonium Facility.

For example, the Laboratory’senvironmental management or-ganization plans to send a repre-sentative to work and reside atthe Plutonium Facility. This rep-resentative will gain an intimateunderstanding of the needs ofthose who generate the waste,and communication will be en-hanced. A communication net-work is also being developed totranslate requirements from thefederal regulations to under-standable directions at the oper-ating level. Monthly meetingsare being held to discuss issuesof mutual interest. Although allpersonnel have full work sched-ules and have difficulty commit-ting time for additional activi-ties, the time required for thesemeetings is well spent.

Operating personnel must beconvinced that waste manage-ment regulations are not trivialand must be observed consis-tently. Waste certification andcharacterization and wastestream control require addi-tional manpower and financialexpenditures. Operating person-nel must clearly understand thesignificance of this additionalwork, even if their principal in-terest is in research. Specifictraining must be developed forall those who generate waste atthe Plutonium Facility. Thetraining program must be wellorganized and must use knowl-edgeable waste managementpersonnel as instructors.

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Matrixelement


Need for Planning andScheduling WasteManagement ActivitiesScheduling waste managementactivities has become essentialbecause the waste handlingprocess has become complex,requiring the support of dozensof personnel. Waste can nolonger be collected and dis-posed of as it is accumulated.Planning must begin before thewaste is generated, and schedul-ing must allow personnel toplan and coordinate their activi-ties. Special activities, such asglove-box removals, must bescheduled to optimize use oflimited waste management re-sources and to avoid conflictswith normal Facility activities.Groups are experiencing staffingshortages, which limit the avail-ability of backup personnel forwaste support activities. Whenunplanned activities do occur,they must be handled with spe-cial care to prevent surprisesthat can derail normal waste dis-posal operations.

The Plutonium Facility WasteManagement TeamThe waste management processat the Plutonium Facility mustbe performed in a cost-effectivemanner while meeting all regu-latory requirements. A wastemanagement team has beenformed to coordinate effortsat TA-55 to attain these goals.The team provides an interfacebetween regulations and opera-tions by providing guidance tothose generating waste. Thisguidance includes how to certifyand package waste according to

established waste acceptance cri-teria. Waste management activi-ties are conducted in a way thatminimizes the impact on opera-tions as much as possible. Wastemanagement personnel must becareful to respect and under-stand the interests of those whogenerate waste and try to avoidrestricting their operations un-necessarily. At times, innovativeways of complying with regula-tions must be implemented.

Shortcomings of the CurrentWaste Management ProgramThe waste management pro-gram must be able to react expe-diently to regulatory changes.For example, the TRU wasteprogram at Los Alamos consistsof a hierarchy of documents.EPA regulations, Departmentof Transportation and Occupa-tional Safety and Health Admin-istration regulations, AEArequirements, and DOE ordersare the highest-level documents.Below them are second-leveldocuments telling how the ap-plicable regulations and orderswill be implemented at disposalfacilities (WIPP) and the Labora-tory. The third and final level ofdocuments are those written atthe Plutonium Facility such aswaste stream attachments tothe Laboratory certificationplans and the Safe OperatingProcedures (SOPs), which de-scribe how work is to be per-formed. Changes to higher-leveldocuments must be reflected inlower-tier documents in a timely

manner, down to the operatinglevel. The revision and reviewprocess for a Plutonium FacilitySOP can easily take a year. Muchof this time is spent on resolvingcomments and questions fromreviewers. With three levels ofprogram documents to update,it could take three years forchanges to filter down to theworking level. This time span isclearly unacceptable. Somethingmust be done to expedite the re-view and approval process if op-erations are to keep current withchanging requirements.

With increasing frequency, thedemand for manpower to keepup with paperwork require-ments is being met by the useof contractors. If contractors areused, it is imperative that they

1) demonstrate regulatory expertise regarding the document that is being written and2) possess knowledge of the operations for which the documents are being written.

The sponsor cannot simply paythe contractor and walk awayfrom the project expecting to re-ceive a satisfactory product. Ifanything, extra attention mustbe given to the development ofcontractor-generated programdocuments. Program documentsmust be written realistically. It isvery easy to make a commit-ment that looks good on paperbut is impossible to carry out.These unrealistic commitmentsare usually the first findings tobe identified during an audit.

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Evolution of Waste Manage-ment at the Plutonium FacilitySeveral projects are currentlyunderway to improve the effi-ciency of the Plutonium Facilitywaste management program.The first is the computer-basedNuclear Materials Technology(NMT) Waste ManagementSystem. This system will takethe place of the current papersystem, which requires stan-dardized forms, manual transferof data, and operator calcula-tions. The hard-copy form usedin the current system must ac-company the waste and func-tions as the data package for thecompleted container. The formmust then be physically circu-lated for review and approval.The new computer-based pro-gram will either reside on oneof the file servers distributedaround the network or on theindividual personal computersconnected to the network andused for data entry in the field.A user-friendly interface willprompt the user for data, checkthe input, do calculations auto-matically, and disallow the entryof improper information. Hard-copy forms for data packageswill be generated outside of theprocessing area, eliminating pa-per transfer across the radiationcontrol areas. Approvals will beelectronic, expediting the reviewprocess. This system should beready for testing in May of 1994.In conjunction with this system,a bar code labeling scheme willbe instituted. Waste items willbe labeled with identification(ID) stickers and scanned asthey are transferred betweenoperations. Data will be enteredunder the item ID and trans-ferred electronically into theNMT Waste ManagementSystem.

The Plutonium Facility wastemanagement team is working toimprove the waste managementsystem at the Facility, not only tomeet the needs of the presentbut also to meet the challengesof the future. Unencumberedcommunication is being encour-aged. For example, personnel atthe Plutonium Facility havebeen issued waste managementtelephone cards to help them getin touch with the right person toassist them. Organizations out-side the facility have been giveninformation about how the Plu-tonium Facility waste manage-ment team is organized. A newfacility-specific training coursefor those who generate waste isbeing planned. The course willbe interactive and will be re-quired for all personnel whocurrently generate waste as wellas new employees. Guidance onprocess knowledge documenta-tion will be emphasized. ThePlutonium Facility is takingadvantage of the DOE MentorProgram, a program by whichDOE offers assistance fromexperts in specialized areas,including environmental com-pliance. These personnel, whoare familiar with regulatory is-sues within the DOE complex,are investigating complianceproblems and are discussingresolution strategies. Their rec-ommendations are currentlybeing evaluated by Facilitymanagement.

SummaryThe Plutonium Facility at LosAlamos is striving to come intofull compliance with all environ-mental regulations for the stor-age, treatment, and disposal ofradioactive and hazardouswaste. The complexity of theregulations paired with theuniqueness of radiochemicalprocessing operations presentschallenges and frustrations tothose involved in the wastemanagement process. Mutuallyagreeable solutions to wastemanagement problems must befound: the EPA and DOE mustcome to agreement on solutionsfor the DOE complex as a whole,and the NMED and Los Alamosmust agree on solutions for thePlutonium Facility in particular.Funding shortfalls and budget-ary constraints are placing agrowing emphasis on minimiz-ing waste generation and maxi-mizing the efficiency of wastemanagement practices. Thewaste management team ismaking steady progress in im-proving the quality of the wastemanagement program at thePlutonium Facility. These im-provements will not only resultin enhanced operational safetyand a reduction in environmen-tal risk but will also ultimatelyresult in full compliance withenvironmental regulations andefficient utilization of taxpayerdollars.✦

Facility Operations


pollution prevention technologies for plutonium processing · 2016-10-21 · pollution prevention...

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