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Distribution Category:General, Miscellaneous, and ProgressReports (Nuclear) (UC-500)
ANL--89/29
DE90 018032
ARGONNE NATIONAL LABORATORY9700 South Cass AvenueArgonne, IL 60439
NUCLEAR TECHNOLOGY PROGRAMSSEMIANNUAL PROGRESS REPORT
October 1987-March 1988
Chemical Technology Division
M. J. Steindler, DirectorJ. E. Harmon, Editor
August 1990
A major purpose of the Techni-cal Information Center is to providethe broadest dissemination possi-ble of information contained inDOE's Research and DevelopmentReports to business, industry, theacademic community, and federal,state and local governments.
Although a small portion of thisreport is not reproducible, it isbeing made available to expeditethe availability of information on theresearch discussed herein.
TABLE OF CONTENTS
Page
ABSTRACT ............................................................. 1
SUMMARY .............................................................. 1
I. APPLIED PHYSICAL CHEMISTRY..................................... 7
A. Studies of Fission Product Release......................... 7
1. Downstream Behavior of Volatile Fission Products ...... 72. Fission Product Release from Core-Concrete Melts ...... 13
B. Thermophysical Properties Studies.......................... 20
1. Thermal Expansion...................................... 202. Solidus and Liquidus Temperatures of
U-19 wt % Pu-10 wt % Zr................................ 223. Fuel-Cladding Interaction Studies...................... 22
C. Tritium Target Development................................. 25
D. Fusion-Related Research.................................... 28
1. Thermodynamics and Kinetics of BreederMaterials.............................................. 28
2. Modeling of Tritium Behavior in CeramicBreeder Materials...................................... 34
3. Design Studies of International ThermonuclearExperimental Reactor................................... 42
4. Tritium Control Experiment............................. 425. Dosimetry and Damage Analysis.......................... 46
REFERENCES ........................................................... 52
II. SEPARATION SCIENCE AND TECHNOLOGY.............................. 55
A. Development of Extraction-Behavior Modelsfor the GTM ............................................... 55
1. Modeling of Nitric Acid Extraction..................... 562. Modeling of Americium Extraction....................... 60
B. Estimating Densities of Complex Aqueous Solutions.......... 64
1. Density Correlation.................................... 642. Density Correlation vs. Measured Density Data ........... 68
C. SASSE Development.......................................... 72
D. Development of the Generic TRUEX Data Base................. 75
E. Measurements of the Extraction Behaviorof Americium and Nitric Acid............................... 75
1. Effects of Nitrate Salts on Am Extraction.............. 752. Nitric Acid Extraction................................. 75
111
TABLE OF CONTENTS (contd)
Page
F. Effects of CMPO and TBP Concentrationon Extraction Behavior..................................... 78
1. Americium Extraction................................... 782. Technetium Extraction.................................. 79
G. Verification Studies....................................... 82
1. Introduction........................................... 822. Future Plans........................................... 823. Verification Run 1..................................... 834. Third-Phase Formation with the
TRUEX-NPH Solvent...................................... 87
H. Centrifugal Contactor Development.......................... 89
1. Contactors for Remote Handling......................... 902. Rotors for TRUEX-NPH Solvent........................... 923. Contactor Operation at Elevated Temperatures........... 934. Minicontactors......................................... 945. Vibration Measurement and Analysis..................... 100
I. Radiolysis and Hydrolysis of TRUEX-NPH Solvent............. 103
1. Introduction .......................................... 1032. Procedure ............................................. 1033. Results ............................................... 104
J. PUREX-TRUEX Processing of Chloride Salt Residues........... 107
1. Introduction .......................................... 1072. Plutonium Model Development............................ 107
K. Production and Separation of 99 Mo from LEU................. 110
1. Introduction .......................................... 1102. Sorption and Desorption of Molybdenum
on Alumina Columns..................................... 1123. Preparation and Irradiation of U3 Si2 . . . . . . . . . . . . . . . . . . . . 1174. Chemical Processing, Separation, and
Analysis of Irradiated U3 Si2 . ........ ....... ...... .. ... 1185. Test of Acid Dissolution for
Unirradiated U3 Si2 -. ... .. .. .. ... .. .. ....... ....... ..... 1246. Future Work ........................................... 124
REFERENCES ........................................................... 124
III. HIGH LEVEL WASTE/REPOSITORY INTERACTIONS ...................... 127
A. NNWSI Unsaturated Test Method.............................. 127
B. NNWSI Parametric Experiments............................... 127
1. Obsidian .............................................. 1.322. ATM-lc ................................................ 136
3. WV 44 ................................................. 137
iv
TABLE OF CONTENTS (contd)
C.
D.
E.
F.
G.
H.
I.
REFERENCES
Relative Humidity and Simple Glass Experiments......
Analytical Support..................................
1. Development of SIMS Analysis Instrumentation ....2. Development of a Laser Raman Microprobe System
Basalt Analog......................................
Vapor Hydration Experiments.........................
Spent Fuel Leach Tests..............................
Radiation Effects...................................
Spent Fuel Literature Review........................
.....................................................
v
Page
140
145
145145
146
146
149
151
154
154
NUCLEAR TECHNOLOGY PROGRAMSSEMIANNUAL PROGRESS REPORTOctober 1987-March 1988
ABSTRACT
This document reports on the work done by the NuclearTechnology Programs of the Chemical Technology Division, ArgonneNational Laboratory, in the period October 1987-March 1988. Work inapplied physical chemistry included investigations into the pro-cesses that control the release and transport of fission productsunder accident-like conditions. the thermophysical properties ofmetal fuel and blanket material of the Integral Fast Reactor, andthe properties of selected materials in environments simulatingthose of fusion energy systems. In the area of separation scienceand technology, the bulk of the effort is concerned with developingand implementing processes for the removal and concentration ofactinides from waste streams contaminated by transuranic elements.Another effort is concerned with examining the feasibility ofsubstituting low-enriched for high-enriched uranium in the pro-duction of fission product *9 Mo. In the area of waste management,investigations are underway on the performance of materials inprojected nuclear repository conditions to provide input to thelicensing of the nation's high-level waste repositories.
SUMMARY
Applied Physical Chemistry
Calculational and experimental efforts are underway to investigatefission product release and transport from a light water reactor (LWR) underaccident conditions. In one such effort, the downstream behavior of fissionproduct I, Te, and Cs is being investigated by injecting Te/Te02 , CsOH, CsI,or mixtures of two or more cf these substances into a flowing superheatedsteam/hydrogen gaseous stream and observing the physical and chemicaltransformations that take place when this mixture is sent down a reaction ducton which a temperature gradient (1000 C down to 200 C) has been imposed.Temperatures were measured under steam flow and external heating conditionssimilar to those used in earlier experiments in which solids were injected.These data have been used to estimate the gas and wall temperatures at one-foot intervals along the duct during earlier experiments. The estimatedtemperatures were found to be significantly different from previous estimates.
Another effort involves investigation into the release of refractoryfission products from the molten core-concrete mixtures that would form if amolten core penetrated the bottom of a reactor vessel in a severe accident.The vaporization of core-concrete mixtures is being measured by the tran-spiration method. In these experiments, mixtures of stainless steel, concrete(limestone or basaltic), and urania (doped with La2 03 , SrO, BaO, and Zr0 2) arevaporized at 2150 K from a zirconia crucible into flowing He-6% H2-0.06% H20.The fraction of the sample vaporized is determined by weight change and bychemical analyses on the condensates that are collected in a condenser tube.
1
2
The results are being used to test the thermodynamic data base and the under-lying assumptions of computer codes used for prediction of release during a
severe accident.
Measurements are being performed to provide needed thermodynamic andtransport property data for Integral Fast Reactor (IFR) fuels. Currentexperiments focus on (1) determination of the thermal expansion of D9 claddingalloy, (2) study of effects of annealing on thermal expansion of fuel,(3) measurement of solidus and liquidus temperatures for U-19 wt% Pu-10 wt%Zr, and (4) differential thermal analysis (DTA) of cladding/fuel mixtures.The DTA tests are of particular interest. These have been performed withmixtures of metal fuel and the cladding alloy HT9 in an effort to clarifyearlier heating experiments that showed evidence of a liquid phase at thefuel-cladding interface when irradiated U-10 wt% Zr clad in HT9 was heated to7500C. No interaction was indicated by the DTA curves at 7000C. A key factorin understanding this result is the existence of a eutectic in the U-Fe systemat 725 C. On the basis of the DTA curves and scanning electron microscopyanalysis of U-Zr/HT9 residues, we believe that on cooling from the liquidstate, most of the Zr precipitated as an intermetaJ.lic compound with Fe,largely Fe2 Zr. On further cooling, Fe2 U precipitated from solution. Finally,at a eutectic temperature, a mixture of Fe2 U and FeU6 formed.
Use of the IFR has been proposed for breeding tritium from lithium oxide.For safety and efficiency, tritium loss to the sodium coolant by permeationthrough the cladding must be less than 1% per year. An effort is thereforeunderway to test methods for limiting tritium loss. In this report period,two copper capsules and one nickel capsule were loaded with Li2 0 and tritiatedwater to test their tritium permeation rates between 310 and 4600C. Thetritium permeation rate in the nickel capsule agreed with the rate in theliterature, but was greater than the 1%/yr limit at 3450C. The permeationloss rate through copper was less than 1%/yr at temperatures below 4300C.
A critical element in the development of a fusion reactor is the blanketfor breeding tritium fuel. Several studies are underway with the objective ofdetermining the feasibility of using lithium-containing ceramics as breedermaterial. Also underway are design studies and dosimetry and damage analysis.
Measured thermodynamic and kinetic data are being related to tritiumretention and release from ceramic tritium breeder materials. Adsorption ofH20(g), dissolution of 0H~, and evolution of H20(g) are being measured at hightemperature for the LiAl02-H2 0(g) system to provide thermodynamic and kineticdata for these processes for LiA102 , a candidate tritium breeder material.Corrections have been applied to our earlier isotherms for H2 0 adsorption andOH- solubility on and in LiA10 2 for the temperatures of 400, 500, and 600C.The adsorption curves indicated lower adsorption at 600C than at 400 or 500'Cbut greater adsorption at 5000C than at 4000C. The observed higher degree ofadsorption at 500 C compared with that at 400 C reflects two chemisorptionprocesses, each with different activation energies. Corrected isotherms forhydroxide solubility in LiA102 indicated decreasing solubility with risingtemperature.
In another effort, we have developed a computer model to predict tritiumrelease f--om a ceramic breeder. This model is based on diffusion and
3
desorption as the rate-controlling mechanisms of tritium release. Experi-mental data for tritium release from the CRITIC experiment (Li2O breeder) atChalk River were unusual in that, when the temperature was increased from 455to 6200C, the tritium release dropped. Attempts to fit the release curves forthese data to our diffusion-desorption model failed. Further analysisindicated that the probable explanation for the observed release curves isthat there are two mechanisms or sites for desorption, each with differentactivation energies. Recent literature data for lithium-containing ceramicssupport this hypothesis. We will use a model with an activation energy ofdesorption that is dependent on surface coverage to determine if such a modelcan reproduce the CRITIC results.
In fusion-related design studies for the international thermonuclearexperimental reactor (ITER) project, we are assessing radiolysis that would beproduced in a 2-1OM aqueous lithium nitrate solution, a blanket designconcept. In addition, experimental efforts are underway to assess tritiumcontrol by oxidation. The management of tritium transport within high-temperature fusion blanket systems is required to achieve an environmentallyacceptable fusion facility.
In neutron dosimetry and damage analysis, neutron facilities are beingcharacterized in terms of neutron flux and energy spectrum, which can be usedto calculate atomic displacements and transmutations. These damage parameterscan also be used to correlate properties changes and to predict materialsperformance in fusion reactors. In this report period, we completed fiveneutron spectral measurements at Argonne's Intense Pulsed Neutron Source(IPNS) using the multiple foil activation technique. The measurements will beused for determining fast neutron effects at IPNS. In other work, neutrondosimetry measurements and radiation damage calculations were completed forthree irradiations in the Omega West Reactor at Los Alamos NationalLaboratory. These irradiations were conducted by Hanford EngineeringDevelopment Laboratory and were designed to compare rdiation damage producedin a fission reactor spectrum with damage produced by 14 MeV neutrons at theRotating Target Neutron Source II at Lawrence Livermore National Laboratory.
Separation Science and Technology
The Division's work in separation science and technology is mainly con-cerned with removing and concentrating actinides from waste streams contam-inated with transuranic (TRU) elements by use of the TRUEX solvent extractionprocess. The extractant found most satisfactory for the TRUEX process isoctyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide, which is abbre-viated CMPO. This extractant is combined with tributyl phosphate (TBP) and adiluent to formulate the TRUEX process solvent. The diluent is typically anormal paraffinic hydrocarbon (NPH) or a nonflammable chlorocarbon such astetrachloroethylene (TCE). Other projects are concerned with (1) developingPUREX-TRUEX flowsheets for extraction of plutonium and other TRU elements fromacidic chloride-containing residues generated by plutonium-metal productionand (2) examining the feasibility of substituting low-enriched uranium for thehigh-enriched uranium currently used in the production of 9*Mo.
TRUEX Technology Base Development. This effort involves development of ageneric data base and modeling capability for the TRUEX solvent extractionprocess. The Generic TRUEX Model (GTM) will be directly useful for site-specific flowsheet development directed to (1) establishing a TRUEX process
4
for specific waste streams, (2) assessing the economic and facility require-ments for installing the process, and (31 improving, monitoring, and con-trolling on-line TRUEX processes. The GiM is composed of three sections thatare linked together and executed by HyperCard and Excel software. The heartof the model is the SASSE (Spreadsheet Algorithm for Stagewise SolventExtraction) code, which calculates multistaged, countercurrent flowsheetsbased on distribution ratios calculated in the SASPE (Spreadsheet Algorithmsfor Speciation and Partitioning Equilibria) section. The third section of theGTM, SPACE (Size of Plant and Cost Estimation), estimates the space and costrequirements for installing a specific TRUEX process in a glove box, shielded-cell, or canyon facility.
Mathematical models of equilibrium and kinetic extraction data for theGTM continue to be improved as more data are collected. In this reportperiod, models were developed for extraction of nitric acid and americium byvarious TRUEX solvents. In addition, a density correlation was derived to beused for converting molar concentrations of solutions to molality in the GTM.
In laboratory studies to obtain data for the GTM, we determined theextraction behavior of nitric acid and americium, the effect of CMPO and TBPconcentration on extraction of americium and technetium, and the effect ofradiolysis and hydrolysis on americium extraction. In addition, a series oflaboratory verification studies has been initiated to develop a betterunderstanding of the chemistry of the TRUEX process, test and verify processmodifications, and verify the results from GTM predictions. In the firstverification run, the extraction of neodymium from a simplified Current AcidWaste solution was investigated using a TRUEX-NPH solvent. Problems wereencountered with formation of a second organic phase, and steps will be takenin future verification runs to resolve this problem.
As part of the TRUEX technology base development project, the design forthe Argonne centrifugal contactor is being modified to meet the needs ofspecific solvent extraction processes. Recently, a 4-cm dia centrifugalcontactor design was modified so that the contactors could be used in a remotefacility. An eight-stage unit was then operated in a shielded-cell mockuparea and a glove box. In both cases, the unit worked as specified. Thehydraulic performance of the unit was tested and altered so that it works asdesigned. In other contactor work, a small unit (2-cm dia) for laboratory useis being designed and built. This contactor will require only about one-tenththe liquid volume required for the 4-cm contactor. In support of contactordevelopment, our methods for measuring and analyzing vibrations are beingimproved.
PUREX-TRUEX Processing of Chloride Salt Residues. In this project,PUREX-TRUEX flowsheets are being developed for extraction of plutonium andother TRU elements from acidic chloride-containing residues generated duringplutonium-metal production. The flowsheet under development includes a TBP-based extraction step for recovering plutonium from pyrochemical salt residuesand a CMPO-based step for making the raffinate nonTRU. In this report period,a model that predicts the Pu(IV) distribution ratio between TBP in TCEsolutions and acidic chloride media was developed. This model was found topredict the Pu(IV) distribution ratio over four orders of magnitude (from 10-2to 102) of hydrochloric acid activities.
5
Production and Separation of 9 9Mo. The effect of low-enriched uranium(LEU) substitution for high-enriched uranium (HEU) in 9 9 Mo production wasstudied. The use of uranium silicide fuels for uranium aluminide requireschanges in current chemical processing, especially during the initial basicdissolution step. The process envisioned now entails dissolution of thecladding in basic solution, followed by dissolution of the uranium silicide ina basic peroxide solution. The peroxide is then destroyed by heating,resulting in the precipitation of uranium. After separation of the uraniumprecipitate, the molybdenum-containing solution is diluted, acidified, andloaded onto an alumina column. The adsorbed 99Mo is eluted with ammoniumhydroxide and further purified. In this report period, the sorption anddesorption conditions of molybdenum on alumina columns were investigated.Also, a U3 Si2 target was prepared and irradiated. This target was thendissolved, the 9 9 Mo was separated from other components on an alumina column,and gamma spectrometry was used to measure the activities of fission productsand neptunium in four sample fractions. Results of these separations arepresented.
High Level Waste/Repository Interactions
The Nevada Nuclear Waste Storage Investigation (NNWSI) Project isinvestigating the tuff beds of Yucca Mountain, Nevada, as a potential locationfor a high-level radioactive waste repository. As part of the waste packagedevelopment portion of this project, which is directed by Lawrence LivermoreNational Laboratory, we are performing experiments to study the behavior ofthe waste form under anticipated repository conditions. This effort includes(1) the development and performance of a test to measure waste form behaviorin unsaturated conditions, (2) the performance of experiments designed tostudy the behavior of waste package components in an irradiated environment,and (3) the performance of experiments to investigate the reaction of glasswith water.
Because the NNWSI Unsaturated Test rigidly sets many of the testparameters, the effect that each parameter may have on the final radionucliderelease needs to be studied. This is being done in parametric experiments. Inthis report period, we conducted hydrothermal leaching (submersion of samplein deionized water) and vapor hydration (sample reacted with water vapor)experiments with ATM-ic and WV 44 glasses and obsidian at 188*C. The obsidiansamples appeared to have undergone some leaching or etching in thehydrothermal experiments. In the vapor hydration experiments, the obsidiansurface appeared unreacted when examined with optical microscopy and scanningelectron microscopy (SEM). With the ATM-1c glass, there was some indicationthat the reaction layer growth follows t/2 kinetics in the hydrothermalexperiments, and the vapor-hydrated ATM-ic showed little evidence forreaction. In the hydrothermal and vapor experiments, the WV 44 glass reactedto form a distinct altered layer penetrating into the glass; the layerthickness increased with time.
Leach tests for 28 days at 90*C were conducted to determine the effect ofnonbridging oxygen ions on the process of hydration with glasses similar tothose used for nuclear waste. The test results are being analyzed.
A laser Raman microprobe (LRM) system is being developed to analyzemicrocrystalline precipitates which form during the hydration of nuclear waste
6
glass. The LRM should be particularly useful in analyzing samples that aretoo small to be removed from the glass surface and studied by X-raydiffraction. Secondary ion mass spectroscopy is being used to profileelements in unreacted glass surfaces.
Hydrothermal leaching and vapor phase hydration experiments have alsobeen performed using two synthetic basalts, SRL 165 glass, and deionizedwater. The reacted glasses are being quantitatively analyzed with respect tothe layer composition and growth kinetics of the layer. Hydrothermal leachingand vapor phase hydration experiments are also in progress with WV 50 glassunder an equilibrium pressure which is above 14 atm at 200 C.
A study has been initiated to determine radionuclide release rates fromspent fuels immersed in site-specific groundwater at ambient hot celltemperature. Leaching tests will be conducted at room temperature in 304Lstainless steel vessels on bare fuel specimens submerged in 250 mL of J-13water. Isotope dilution techniques will be used to measure the dissolutionrate of U02 in water. A review of the spent fuel literature is being done toidentify, collect, and evaluate existing literature on the performance andcharacteristics of spent fuel under conditions relevant to a high-levelnuclear waste repository.
Determining the effects of temperature and gamma radiation on thegaseous/aqueous environment of the high-level waste package is important incharacterizing its performance during the containment period. Experiments arein progress to determine the NO, yield as a function of temperature andabsorbed dose in 304L stainless steel vessels under conditions expected in atuff environment. Initial experiments were done with dry air at 28*C,adsorbed doses of 4.3-308 Mrad and dose rates of 0.1-0.4 Mrad/h, and testperiods of 1 day to 3 months. The yield of nitrous oxide was found toincrease linearly with an increase in absorbed dose. In addition, there was aninverse correlation between the yields of CO2 and NOR. Unusually low NO.yields were accompanied by the formation of a relatively high amount of CO2 .
7
I. APPLIED PHYSICAL CHEMISTRY(C. E. Johnson)
The program in applied physical chemistry involves studies of the thermo-chemical, thermophysical, and transport behavior of selected materials inenvironments simulating those of fission and fusion energy systems.
A. Studies of Fission Product Release
The objective of this effort is to understand the release and transportof fission products from light water reactors under accident-like conditions.
1. Downstream Behavior of Volatile Fission Products
(I. Johnson)
In previous work, the downstream behavior of fission product I, Te,and Cs was investigated by injecting Te/Te02 , CsOH, CsI, or mixtures of twoor more of these substances into a flowing steam/hydrogen environment. Thegaseous mixture was subjected to a temperature gradient as it flowed down a12-ft (3.6-m) long reaction duct. The reaction duct was lined with 12 remo-vable liners. The changes in the gaseous system were inferred from analysesof material deposited on the walls of the reaction duct. These studies areexpected to aid in predicting the behavior of volatile fission products in theprimary system during loss-of-coolant accidents.
Additional thermocouples were installed in the duct to measure thetemperature of the flowing gas stream and the inner wall. Five experimentshave been performed under duct heating and gas flow conditions similar tothose used in the earlier experiments except that steam was injected ratherthan a fission product solution. The results of these experimnts were usedto estimate the temperatures in the duct during the experiments in whichfission product samples were introduced. The previously used temperatureswere calculated using a model of the system in which the steam was assumed tohave the same temperature as that measured at the entrance to the duct andthen to be cooled (as it flowed down the duct) by the transfer of heat byradiation and convection to the wall of the duct, which was assumed to havethe same temperature as the outside surface. A steady state was assumed.
The results of the present experiments indicate that the system isnot at a steady state, i.e., the temperature is continuously changing. Thevariation of the temperature profile is due to the flow of heat into the ductfrom the external heaters, axially down the duct due to thermal conductance,and to the flow of hot gas down the duct. The inner duct liners are not atthe same temperature as the outer wall of the duct. The temperature of aliner is determined by the transfer of heat from the hot flowing gas and theouter wall. Heat may also flow axially in the liners. This leads to adynamic condition in which the temperature is dependent not only on the gasflow and power into the external heaters but also the time.
The temperatures observed for the gas, the inner wall of the liner,and the outer wall of the duct for Expt. 11-1 are shown in Fig. I-1. Similarresults were obtained in the four other experiments. It is seen that the
8
1000 /-OUTSIDE PIPE
o 800
STEAM
600
NNER WALL s '-.400
2000 2 4 1 8 10 12
LOCATION, ft
Fig. I-1. Estimated Temperatures Along Duct forthe Gas, Inner Wall, and Outside Wallduring Expt. 11-1
temperature of the gas and the inner wall at the entrance to the duct is lowerthan the temperature on the outside of the duct. At about four feet (1.2 m)down the duct, the temperature of the inner wall of the liner and the temper-ature of the outer wall of the duct become equal. Beyond six feet (1.8 m),the inner wall temperature is greater than the temperature of the outer duct.There is also a cross-over point, at three feet (0.9 m), where the temperatureof the steam becomes greater than the temperature of the outer wall. Theseobserved temperatures for the steam and the inner wall are significantlydifferent from what had been previously estimated. The steam temperature ismuch lower than previously estimated.
The lower temperature of the steam and the liner at the entrance tothe duct compared with that of the outside of the duct is caused by a coolingof the steam as it is transported from the superheater to the duct. (Theinjection section is located between a superheater and the duct.) The injec-tion section and connecting pipe are 23-in. (158-cm) ling. Only about 6 in.(15 cm) of this connecting section is heated by external heaters. The con-necting section is insulated with 4-6 in. (10-15 cm) of fiberfrax. During theheatup of the system, prior to the sample injection, this connecting sectionis heated by transfer of heat from the flow of steam and by heat conductancefrom the superheater and the duct. The sample was injected after the observedtemperature in the steam, as measured by a thermocouple in a well adjacent tothe external heaters, had reached a constant temperature. This thermocouplewas about 13 in. (33 cm) from the end of the duct. The steam cooled as itflowed from this point to the entrance of the duct. The amount of coolingdepended on the temperature of the connecting pipe, which, in turn, dependedon how long the system had been heated prior to injection.
The steam temperatures are plotted against the outer duct tempera-ture in Fig. 1-2 for four runs. A linear relation was used to correlate the
9
1000
900
800
700
600
500
400
300I i
-
00
-p
- v Q11-1
0 11-2- o 11-3
v o0 V 11-4"
0 11-5* NOT IN DATAUSED FORUNES SHOWN
I II
0000 400 60') 800TEMPERATURE OUTER WALL, C
Fig. 1-2.
1200
Estimated Temperatures of Steam vs.Outside Duct Temperatures
steam and the outer wall temperatures. The inner wall temperature was alsocorrelated with the outer duct temperature by a linear relation. These linearrelations were used to estimate the gas and inner wall temperatures for theearlier experiments in which fission product samples were injected. Theestimated temperatures are shown in Tables I-1 to -8. Since the outer walltemperatures were only measured at 2-ft (0.6-m) intervals, the intermediatetemperatures were computed using a cubic spline method. In previous reported
Table I-1. Estimated Temperatures for Expt. 1 (CsI injection)
Temp., *C
Location, Inner Outerft Gas Wall Wall
0 895 893 10221 867 841 9762 837 816 926
3 805 800 8744 774 775 823
5 744 729 774
6 715 6:9 7277 687 643 681
8 659 615 635
9 631 584 588
10 604 553 545
0
D
-
I 1 I l
rA
vvv
I
10
Table 1-2. Estimated Temperatures for Expt. 2 (CsI injection)
Temp., 0C
Location, Inner Outerft Gas Wall Wal]
0 886 871 1006
1 820 810 912
2 779 751 832
3 735 701 760
4 691 652 688
5 646 600 613
6 603 551 543
7 568 511 486
8 539 478 438
9 513 448 394
10 489 422 356
Table 1-3. Estimated Temperatures for Expt. 3 (CsOH injection)
Temp., 0C
Location, Inner Outerft Gas Wall Wall
0 885 871 1005
1 895 888 978
2 851 834 950
3 777 749 826
4 700 661 702
5 639 592 624
6 595 542 529
7 561 504 480
8 528 466 420
9 489 422 360
10 451 378 293
11 423 346 238
12 414 337 233
11
Table 1-4. Estimated Temperatures for Expt. 4 (CsI andCsOH injection)
Temp., 0C
Location, Inner Outerft Gas Wall Wall
0 884 884 1003
1 905 898 1038
2 857 840 960
3 774 745 824
4 690 650 686
5 630 582 587
6 585 531 513
7 543 483 444
8 505 440 382
9 474 404 331
10 447 374 287
11 422 346 246
12 395 316 202
Table I-5. Estimated Temperatures forTe injection)
Expt. 7 (CsI, CsOH, and
Temp., 0C
Location, Inner Outerft Gas Wall Wall
0 883 883 1002
1 892 883 1017
2 833 ¬12 920
3 743 710 7734 662 618 640
5 616 566 565
6 589 535 520
7 559 501 470
8 527 465 418
9 499 433 372
10 470 400 324
11 431 356 261
12 375 293 169
12
Table 1-6. Estimated Temperatures for Expt. 8Te injection)
(CsI, CsOH, and
Temp., C
Location,ft
01
234
5
67
89
10
11
12
Gas
885
877
772
633525
490
487
469444
427
414
396365
InnerWall
871
862743
586
463
422
419
399371
352337
317
282
OuterWall
1005
992
820
593
415
357
352
323
282
254
233
204
153
Table 1-7. Estimated Temperatures for Expt. 9 (CsI, CsOH, andTe injection)
Temp., 0C
Location, Inner Outerft Gas Wall Wall
0 886 872 10071 882 869 10022 798 773 8633 682 642 6734 585 531 513
5 541 481 441
6 523 461 4127 500 435 3748 474 405 3319 455 383 299
10 440 366 27511 426 350 25212 407 329 222
13
Table 1-8. Estimated Temperatures for Expt. 10 (Te injection)
Temp., OC
Location, Inner Outerft Gas Wall Wall
0 882 868 1001
1 842 823 935
2 760 729 800
3 666 623 647
4 593 540 527
5 560 504 4746 543 484 445
7 513 449 395
8 477 408 336
9 450 377 291
10 431 356 260
11 418 342 239
12 408 330 223
temperatures, the intermediate temperatures were computed assuming a linearrelation. In most cases, the differences in the temperatures obtained withthe two methods are not large. The temperatures were estimated for the systemjust prior to the sample injection. In a future report, the temperaturesafter the injection, which are important for understanding the depositiondata, will be estimated. This will be particularly important for Expt. 9,in which revaporization was studied.
2. Fission Product Release from Core-Concrete Melts(M. F. Roche, J. L. Settle, and L. Leibowitz)
In a severe nuclear reactor accident, the core can melt, penetratethe bottom of the reactor vessel, and react with the concrete basemat beneaththe vessel. Our objectives are to measure and calculate the vaporization ofcertain refractory fission products--strontium, barium, and lanthanum--fromthe molten core-concrete mixtures that would form under these conditions. Thevaporization measurements are being done using a transpiration method; thecalculations employ the SOLGASMIX computer code.' Our version of the codecalculates equilibrium conditions for 17 elements in the gas, liquid, andsolid phases. One hundred gas-phase species (elements, oxides, hydroxides)are included in the calculations; there are also 74 solid or liquid speciesincluded in the metal and oxide phases. The thermodynamic data for thesespecies were obtained from a variety of sources.2-7
The amounts of materials employed in the experiments with zirconiacrucibles were limestone-aggregate concrete, 3 g; stainless steel, 3 g; andU02 (doped with 1 mol % La2 03 , 0.4 mol % BaO, 1 mol % SrO, and 2 mol % Zr02),3 g. The same amounts were employed in experiments with molybdenum crucibles
14
except for omitting the steel. The materials were heated at 2150 K within aMo-30W furnace tube (46-cm long, 2.9-cm OD, 2.4-cm ID). The gas flowing overthe sample was He-6% H2-0.06% H20; this H2-to-H20 ratio is equivalent to apartial molar free energy of -420 kJ per mole of 02. The flow rate of the gaswas either 100 or 200 cm /min (measured at 25 C), and about 0.6 mol of gas wasused. The water concentration in the gas was measured with a thin-filmalumina hygrometer (Model 550, Panametrics, Inc.).
Materials vaporized from the sample were collected in a molybdenumcondenser tube (54-cm long, 0.8-cm OD, 0.5-cm ID) whose entrance lay withinthe crucible a few centimeters above the sample surface. Following the run,the condensate was extracted from the condenser tube with acid washes(HC1, HNO3), and the solutions were analyzed by inductively coupled plasma/atomic emission spectroscopy (ICP/AES)* or fluorescence spectroscopy (foruranium).** In addition, the starting materials were all assayed by thesame techniques.
The sample from one of the runs employing a zirconia crucible wascross-sectioned, mounted, polished, and examined by electron probe micro-analysis (EPM).***
Phases. The zirconia crucible from a 146-min run (gas flow rate of100 cm3/min) was examined after cross sectioning. The solidified mass ofconcrete, urania, and steel, which occupied the bottom third of the crucible,had evidently been at least partially liquid during the run. The concrete-urania phase, which had wet the zirconia crucible, had a pronounced meniscus.In addition, the thickness of the zirconia-crucible wall had been reduced byroughly 40% in the wetted area (the bottom third of the crucible). The steelphase (approximately spherical) was embedded within the solidified concrete-urania mixture; evidently the steel phase could not readily equilibrate withthe gas phase in this geometry.
The EPM results for two areas of this sample are shown in Figs. 1-3and -4. A secondary electron image and maps of the distribution of the majorelements (U, Zr, Ca, Mg, Si) are shown for each area. Also detected by EPM inspot analyses were the elements Al, La, Ba, Sr, Cr, and Fe. The La and Srdistributions tended to parallel the Ca distribution. The signal for Ba wastoo weak to draw any general conclusions about its distribution, but it wasdetectable in several regions high in uranium.
The major phases present in the mixture were inferred from theoverlap of the elemental images in the photomicrographs. The phases were(1) urania containing calcia and zirconia, (2) calcium zirconate, (3) acalcium-magnesium silicate, and (4) magnesia associated with the calciummagnesium silicate. About 10% of the zirconia crucible was dissolved by theconcrete-urania mixture during the experiment, which accounts for the presenceof zirconia-containing major phases. The urania and calcium-zirconate phasesseen in the sample have melting points above 2500 K and were probably solid
*Performed by E.A. Huff and K. J. Jensen, ANL Analytical Chemistry
Laboratory.
**Performed by A. Essling, ANL Analytical Chemistry Laboratory.
***Performed by C. A. Seils, CMT Division.
15
Secondary Electron 100OX
I A4N r
, ., a .4
Zirconium 1000X
Magnesium 1000X
Fig. I-3.
Uranium 1000X
Calcium 1000X
Silicon 1000X
Electron Probe Microanalysis of 100 x 80 pm Areaof Sample from 146-min Run at 2150 K. The samplecontained limestone-aggregate concrete, dopedurania, and stainless steel.
L .10tid$
16
Secondary Electron 1000X
Zirconium 1000X
Uranium 1000X
Calcium 1000X
Magnesium 1000X Silicon 1000X
Fig. I-4. Electron Probe Microanalysis of Second 100 x 80 14mArea of Sample from 146-min Run at 2150 K. Thesample contained limestone-aggregate concrete,doped urania, and stainless steel.
17
during the experiment. The calcium-magnesium silicate and magnesia phasesprobably crystallized from a liquid phase. This phase had a composition neara eutectic (mp, 2070 K) on the calcium orthosilicate-magnesia tie line in theCaO-MgO-Si02 phase diagram,8 according to SOLGASMIX calculations (see below).
Assays. The analytical data from the runs in the zirconia crucibleshave been presented elsewhere,9 but the detailed EPM results on system phaseswere not available at that time. In the runs, the amount of material trans-ported at a gas flow rate of 100 and 200 cm3/min was nearly the same (98 and129 mg, respectively), indicating that gas saturation had been achieved. Theanalytical-chemistry data from the runs are shown in the last two columns ofTable 1-9, while the amounts added to the beaker are listed (as elementalmasses) in the second column, and the SOLGASMIX calculation is listed in thethird column. The "Amounts Added" column includes our estimate of zirconiadissolution from the beaker and is a factor of 100 higher than the amountincorJrated as a dopant in the urania.
A comparison of the two columns of analytical data in Table 1-9suggests that the major limestone-concrete components, Mg and Ca, had achievedsaturation because the amount of transport was independent of gas flow rate.However, we know that the major steel components Fe, Cr, and Ni did not satur-ate the gas phase. They exhibited much less transport in these runs than in
Table I-9. Calculated and Measured Vapor Transport in Runs with ZirconiaCrucibles Containing Stainless Steel, Limestone-AggregateConcrete, and Doped Urania (2150 K)
Measured Transport,a mgAmount Calculated
Element Added, mg Transport, mg 200 cm3/min 100 cm3 /min
La 33 0.015 <0.002 (0.005
Ba 6 0.019 0.018 0.028
Sr 12 0.010 0.006 0.012
U 2648 0.087 0.047 (0 .0 0 5)b 0.028 (0.014)Fe 2470 44 1.26 (1.00) 1.52 (0.91)
Cr 628 33 1.64 (0.97) 1.36 (0.54)
Ni 277 3 0.08 (0.06) 0.04 (0.02)
Mn 50 38 15.4 (3.4) 9.7 (3.6)
Zr 2056 0 0.009 <0.005
Ca 974 0.6 0.2 0.22
Mg 143 67 40 39
Si 118 1.8 2.9 ----Al 32 31 0 0.2
'Analyses of the first HCl etch and thecondenser tubes were combined.
second (more severe) HNO3 etch of Mo
bValues in parentheses are analyses on the second (more severe) HNO3 etch ofthe Mo condenser tube. Note diffusion of some elements (mainly Fe, Cr, Ni,Mn, but also U) into the Mo surface.
18
runs with the steel alone, in which gas saturation by the steel components wasdemonstrated by both analytical data and SOLGASMIX calculations.9 The lack ofsaturation in the present runs is attributed to the steel phase being embeddedin the concrete-urania phase.
The data from the runs in molybdenum crucibles are given inTable I-10. Again, the masses transpired, 86 and 91 mg, respectively, at 100and 200 cm3/min gas flow rate were nearly identical. Also, the amount ofmagnesium (the major element in the deposits) being transported was nearly thesame in both runs.
Table I-10. Calculated and Measured Vapor Transport in Runs withMolybdenum Crucibles Containing Limestone-AggregateConcrete and Doped Urania (2150 K)
Measured Transport,a mgAmount Calculated
Element Added, mg Transport, mg 200 cm3 /min 100 cM3/min
La 32 0.016 0.005 0.003
Ba 6 0.348 0.065 0.033
Sr 12 0.199 0.040 0.031
U 2542 0.100 0.170 (Q.015)b -
Fe 18 18 4.8 (2.4) 10.8 (8.1)
Cr 0.13 0.13 0.02 0.11 (0.10)
Ni 0 0 0 0.02 (0.01)Mn 0.35 0.33 2.3 (0.5) 1.7 (1.1)
Zr 20 0 0 0
Ca 988 1.5 4.6 (0.1) 1.8 (0.2)
Mg 144 70 58 (0.1) 57 (2.5)
Si 107 0.34 --- A---Al 32 32 0 0
aSum of analyses on first HCl etch and second (more severe) HN03Mo condenser tube.
bValues in parentheses are analyses on second (more severe) HNO3Mo condenser tube. Note that a manufacturing error led to somemanganese contamination in the two Mo tubes used in these runs.
etch of the
etch of theiron and
Note that the major changes in the "Amount Added" column betweenTables 1-9 and -10 are in the amounts of the steel components and thezirconium. The measured transport of the alkaline earths is generally higherwith the steel and zirconia at low levels (Table I-10). For barium, it isabout a factor of two higher, and for strontium it is about a factor of fourhigher. The lanthanum also is higher in that it is at least detectable in thevapor condensates for the runs reported in Table I-10. Clearly, these changesare not due to the steel, which forms a separate phase. It is the zirconiathat suppresses vaporization of the alkaline earth oxides by reaction with
19
them to form a mixed zirconate. The decrease in lanthanum vaporization canalso be explained b zirconate formation; the compound La2Zr2 07 (mp, 2573 K)is known to exist.'
SOLGASMIX. The calculated transport shown in Tables 1-9 and -10 isbased on the phase information obtained from the EPM. Previously, we hadassumed that all the solid oxides (including the zirconates of the alkalineearths) formed an ideal solid solution.9 Using the EPM phase data as a guide,we now split the solid oxides into a urania solid solution, an alkaline-earthzirconate solid solution (unfortunately, no thermodynamic data are availablefor lanthanum zirconate), and a "silicate" solid solution. As a result ofincorporating these changes in SOLGASMIX, the measured data and the calcula-tions in Table 1-9 agree reasonably well except for the aluminum and steelelements (Fe, Cr, Ni, Mn). As noted earlier, the steel forms an embeddedphase, and its elements are apparently not transported. The reason for thelack of agreement with aluminum is still under investigation.
If we repeat the calculations for Table 1-9, but with the speciesSrZrO3 and BaZrO3 omitted, the calculated transports of Sr and Ba are 0.167and 0.292 mg, respectively. From the ratio of the calculated transport withand without the zirconates, we derive an apparent activity coefficient ofabout 0.06 for both strontia and baria in the high-zirconia system.
The calculations are less satisfactory for the case where thezirconia concentration is small (Table I-10). Here, the calculated values,which are based on an ideal solution of strontia, baria, lanthana, and urania,are too high by a factor of about five. The problem is that no thermodynamicdata for the silicates of strontium, barium, and lanthanum exist. Includingthese silicates in the data base would no doubt improve the calculations; thesilicate data for magnesium and calcium are included in SOLGASMIX, and theirrelease is predicted quite well. From the ratio of the calculation andmeasurement in Table I-10, we derive apparent activity coefficients forstrontia, baria, and lanthana of about 0.2 in the low-zirconia system.
A continuing problem with the SOLGASMIX calculation is the predic-tion of a solid silicate slag rather than a liquid. We will need to improvethe treatment of the liquid oxides to model the region near the melting pointof these complex silicate mixtures. For example, the calculated silicatephase for the cases presented in Tables 1-9 and -10 is a nearly equimolarsolid solution of calcium orthosilicate and magnesia. However, according tothe MgO-CaO-SiO2 phase diagram,8 a mixture of these solids does not form asolid solution and, in fact, is some 75 K above the melting point. Except forthese problems with the silicate phase, the phases derived in the code are nowin good agreement with experiments.
The release of the refractory fission products (Sr, Ba, La) fromcore-concrete mixtures at 2150 K is much less than is predicted by simplemodels that employ ideal solutions of strontia, baria, and lanthana. Thosesimple models yield release fractions that are e factor of 16 too high. It isnecessary to include the thermodynamics of the zirconates of the alkalineearths and the rare earths to reflect the actual phases in a core-concretemixture.
20
Even in the absence of the zirconates, the amount of vaporization isless than predicted by the ideal-solution model. We believe that thermody-namic data for the silicates of the alkaline earths and rare earths must beincluded to correctly predict the release under these conditions.
The transpiration method provides an effective test of the thermo-dynamic data base and the model assumptions used in the computer codes. Ifthe experiments are extended to other types of concrete, other oxygen poten-tials, and ether temperatures, then reliable predictions can be made for avariety of reactors.
B. Thermophysical Properties Studies(L. Leibowitz, E. Veleckis, and R. A. Blomquist)
Development of the Integral Fast Reactor (IFR) requires understanding offuel behavior for wide temperature and composition ranges. Much of the neededinformation is not available, and o'ir ongoing program is designed to providethe essential thermophysical proper'ky data. These efforts are coordinatedwith R&D efforts on fuel performance, design, and safety to be certain thatfuel properties of primary concern are examined. Our effort involves experi-mental and calculated work.
1. Thermal Expansion
The thermal expansion of cladding alloys (D9 and HT9) is beingmeasured to provide a firmer foundation for these data. This work is per-formed with a Netzsch Model 402 dilatometer installed in a helium-atmosphereglove box. National Bureau of Standards (NBS) reference materials are usedperiodically to calibrate the system.
Thermal expansion data are presented here for D9 cladding alloy.Measurements were performed at heating/cooling rates of 1 K/min up to 1273 K(10000C). No significant hysteresis was observed, and pooled data from fiveheating/cooling cycles are presented in Table I-11. The standard deviation isabout 0.2% and the absolute accuracy is about 1%.
Table I-11. Thermal Expansion of D9Cladding Alloy
Temp., ExpansionC (AL/L), %
100 0.145
200 0.325
300 0.513
400 0.709
500 0.909
600 1.119
700 1.330
800 1.543
900 1.758
1000 1.974
21
These data are reproduced by the following equation with a standarddeviation of about 0.13%:
AL/Lo = -0.024 + 1.638 x 10-3T + 5.667 x 10-7T2 - 2.069 x 10-10T3 (I-1)
where the thermal expansion, AL/Lo, is in percent, and the temperature, T, isin C.
The thermal expansion of U-8 wt % Pu-10 wt % -Zr was also measured,and the results showed significant variation with a number of thermal expan-sion cycles. The results of three thermal expansion cycles are shown inFig. I-5. These measurements are performed at heating/cooling rates of1 K/min. It seems clear that changes are occurring in the high-temperature
100 200 300 400 500 600 700 800 900
Temperature, *C
Fig. 1--5.
Thermal Expansion of U-8 wt %Pu-10 wt % Zr
1000
(7) phase of the alloy. Similar effects have been noted in past work, but inthis case we were careful to follow the progress of the chanes. Previouslyreported thermal expansion data were for stabilized alloys.' Between runs 6and 7 shown in Fig. 1-5, the sample was held at 10150C for 100 h. Thislengthy anneal clearly produced additional changes, and further anneals andthermal expansion measurements are in progress. In addition to the shifts inthermal expansion, the room-temperature length of the sample increased byabout 0.7% from the as-cast condition to just prior to run 7. At the con-clusion of these tests, the sample will be sent for microscopic examinationto discover what structural changes took place to produce these changes.
0
E- ..
t.
- o
-
N
cl -T
av
N
N
O
!V
22
2. Solidus and Liquidus Temperatures of U-19 wt % Pu-10 wt % Zr
Measurements are continuing to determine solidus and liquidustemperatures of IFR fuel alloys. We report here on measurements forU-19 wt % Pu-10 wt % Zr. Solidus and liquidus temperatures were determined bydifferential thermal analysis (DTA) for heating/cooling cycles at ratesranging from 5 to 20 K/min. Generally, solidus temperatures were determinedfrom heating curves, and liquidus temperatures were determined from coolingcurves. At the lower rates, it is expected that departures from equilibriumwould be smaller and more precise values would be obtained. The experimentalDTA peaks, however, were smaller and more difficult to detect. At the higherrates, larger but possibly less precise peaks were obtained. For the presentcase, the values obtained at 5 K/min appear to be unreliable, but from 7.5 to20 K/min consistent solidus and liquidus temperatures were found and no trendswith rate were observed. The measured solidus and liquidus temperatures withtheir standard deviations were 1087 70C and 1305 11 C, respectively.Solidus and liquidus temperatures calculated from the revised Pelton model12were 1075 and 12950C, respectively, which agree with the measurements. Itshould be noted that the composition used is the nominal composition, andthese calculated values will be revised when composition data are obtainedwith scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS).
3. Fuel-Cladding Interaction Studies
Experiments were performed on two fuel-cladding mixtures: (1) 1.39 gof U-8.6 wt % Zr and 0.44 g of HT9 and (2) 2.45 g of U-26 wt % Pu-10 wt % Zr(nominal composition) and 0.39 g of HT9. Each fuel-cladding mixture washeated at 10 K/min to about 15500C in a yttria crucible and held for twohours. Following this treatment the cooling curve was determined for eachmixture. Several additional heating/cooling cycles were obtained with theU-Zr test. Instrumentation difficulties forced us to terminate the U-Pu-Zrtest after the initial heating/cooling cycle.
Similar results were obtained with the two fuel-cladding mixtures.Initial heating, shown in Fig. 1-6 for U-Zr and Fig. 1-7 for U-Pu-Zr, revealedthe expected solid-state transitions for the fuel, not well resolved for theternary fuel in the range 600-700 C. A small indication of a claddingtransition appears at about 9000C for the binary fuel. At 1155C for thebinary fuel and 1105 C for the ternary fuel, however, a large exothermal peakappears, indicating an energetic reaction producing more stable products.This reaction begins slightly below, and overlaps with, the solidus tempera-tures for the fuels. In the binary fuel, we see indications of additionalpeaks at about 1500 C, possibly caused by melting of HT9. On cooling (shownin Fig. 1-8 for the binary and Fig. 1-9 for the ternary fuel), the two fuel-cladding mixtures behave similarly. Only two peaks appear. A fairly broadpeak, which suggests a liquidus, is seen at 12200C for the binary and 1082Cfor the ternary fuel. At about 7000C for the binary and about 6300C for theternary fuel, there is a very sharp transition, which appears to be eutecticfreezing.
R. V. Strain (EBR-II) is examining the fuel-cladding residues fromthese tests. Results have been obtained for the binary fuel test (U-10% Zrand HT9) and indicate the presence of four phases. The first phase showedonly Zr in SEM analysis, although carbon, oxygen, or nitrogen may be present.
23
0;
Fig. 1-6.
Init ial Heating of U-8.6 wt % Zrand HT9
O0
b
0
o'
200 400 600 8 00 1000
Temperature, 0C
200 400 600 800 1000 1200
Temperature, *C1400 1600
Fig. I-?.
Initial Heating of U-26 wt %Pu-10 wt % Zr and HT9
1200 1400 1600
>0o
1 '
.
*
L
24
021
Fig. 1-8.
Cooling of U-8.6 wt % Zr and HT9
200 400 600 600 1000Temperature, *C
E
c-
0 200 400 600 600 1000 1200 1400 1600Temperature, *C
Fig. 1-9
Cooling of U-26 wt % Pu-10 wt % Zrand HT9
1200 1400
F-
SN-
0 '
0
ri
---- '----
o.
h-
o
-
o,
C.;'
1800
25
This corresponds to the so-called globular phase seen in all IFR fuels. Thesecond phase contained about 66 at. % Fe, 13 at. % Cr, 14 at. % Zr, and7 at. % U. This composition corresponds to Fe3 Zr with some solubility foruranium. It is the innermost phase and was probably the first to solidify.The third phase contained 70 at. % Fe, 10 at. % Cr, 16 at. % U, and4 at. % Zr. This phase appears to correspond to Fe2U with some solubilityfor zirconium. The final phase was a continuous matrix with the composition75 at. % U and 25 at. % Fe. These SEM results combined with the DTA findingssuggest the general sequence of events which occurred in this experiment. Insimplified terms, on cooling from the liquid state, most of the zirconiumprecipitated as an intermetallic compound with iron, largely Fe2Zr. Onfurther cooling, Fe2U precipitated from solution. Finally, at a eutectictemperature, a mixture of Fe2U and FeU6 formed. These results show that,while zirconium will react preferentially with iron, the iron-uranium eutectictemperature remains unchanged because of the very low solubility of Zr inalpha-uranium. Additional compatibility tests of this kind have been sus-pended pending SEM examination of the ternary fuel-cladding residue.
C. Tritium Target Development(P. E. Blackburn and D. V. Steidl)
Use of the IFR breeder reactor has been proposed for breeding tritium
from lithium oxide by the reaction,
6 Li2 O + n = 6 LiOT + He (1-2)
where LiOT is in equilibrium with T20 by the reaction,
2 LiOT = T2 0(g) + Li2 0 (1-3)
and T2 0(g) is in equilibrium with tritium gas by
T 2 0(g) = T 2 (g) + 1/2 0 2 (g) (1-4)
In this process 1% of the lithium is converted to tritium per year. Forsafety and efficiency it is necessary to limit tritium loss to the sodiumcoolant by permeation through the cladding to less than 1% per year. Thepurpose of this effort is to test methods for limiting the tritium loss and toidentify materials and configurations to achieve this end.
Three general methods are proposed to limit permeation loss: (1) tritiumgetters, (2) tritium barriers, and (3) tritium oxidation. The permeation rateis proportional to the differences in the square roots of the tritium pres-sures inside and outside the cladding.
Several getters are being considered to lower the tritium pressure insidethe cladding. Yttrium is the most effective in lowering the tritium pressure.Misch metal (a mixture of rare earths) is less effective than yttrium, butless expensive. The SAES getter (an alloy of zirconium) is also less effec-tive than yttrium, but has the advantage of lettering tritium in the presenceof oxygen. The second method involves the use of a tritium barrier with alower permeation rate than that of stainless steel. Tungsten, silica, andceramic-coated cladding have such low permeation rates. The third method is
26
retention of tritium as T20. If the oxygen pressure is high enough, reac-tion 1-4 will be reversed to oxidize tritium to T20, which, in turn, reversesreaction 1-3 to form LiOT. The increased oxygen pressure can be achieved byusing metals whose oxides are relatively unstable. Such metals includeplatinum, gold, silver, and copper.
To investigate these materials, we are sealing Li20 with a small amountof tritiated water in capsules of the material to be studied. The capsule isheated in a stream of helium, which carries the permeated hydrogen and tritiumthrough an oxidizer to convert the hydrogen and tritium to H20, HTO, and T20.These gases are trapped in ethylene glycol bubblers, which are sampled andcounted to measure the rate of permeation. The helium gas includes 1% watervapor to exchange with adsorbed tritium as HTO and T20. This assures completetransport of the tritium to the bubblers. The oxidizer is a mixture of copperand copper oxide.
Our experimental apparatus contains three sets of quartz furnace tubeswith flow meters and two bubblers in sequence to assure complete trapping ofthe tritium. Each of the three furnaces has its own temperature controller.A multipoint recorder displays the temperature of each furnace measured usingchromel-alumel thermocouples.
We found that the reaction of water with Li20 is very slow at low tem-peratures. A temperature of 300 C was necessary for reaction to occur at areasonable rate. There was a potential problem of water loss because of thehigh temperature during sealing of the capsule by welding. Consequently, wedeveloped a procedure to avoid loss of the tritiated water during welding.Capsules with one end open were loaded with lithia, and 6 L tritiated water(100 pCi/cm3) was added with a micropipette. Then the capsule was placed in acopper chill block that had been cooled with liquid nitrogen. This froze thetritiated water inside the capsule. The top of the capsule was crimped shut,and the capsule and chill block were moved to a glove box where the crimpedend was welded closed.
In our first tests we used two copper capsules (designated M and T) andone nickel capsule. The nickel capsule was chosen as a standard becausehydrogen permeation rates are relatively high and well known for nickel.Because the capsules will contain helium at 1 atm at room temperature (or atlower temperature from the chill block), the pressure at test temperatureswill be 2-3 atm. The yield strength of copper at 3 atm is satisfactory atabout 205 C (400 F). To avoid problems at higher temperatures we designed thecapsules so that they could be partially evacuated while being held in a chillblock near liquid nitrogen temperature. The capsules contained a 1/4-in. tubeextension for evacuation and crimping. We established that crimping resultedin a leak-tight seal. However, when the sealed capsules were welded, thecrimped seal was opened and raised the pressure to 1 atm. Because of timeconstraints, we tested all three capsules rather than waiting for fabricationof copper capsules with longer tube extensions to allow double crimping.
The results of our tests are given in Table 1-12. One of the copperexperiments (sample No. 2) was erratic because of poor temperature control andproblems with the thermocouples. New thermocouples were ordered. All permea-tion data were normalized to 1.51-cm2 wall area and 0.038-cm wall thickness
per gram of Li2 0. The normalized nickel results agree well with literaturepermeation measurements.13 The tritium loss from the nickel capsule was 5.7%
27
Table 1-12. Tritium Loss from Nickel and Copper Capsules
Tritium LossSample Time, Temp.,
Material Number days C pCi %/yr
Copper
T 1 11.96 310 0.0540 0.113T 2 2.90 310 -0.0055 -0.047T 3 5.03 310 0.0094 0.047T 4 2.10 334 0.0077 0.092T 5 4.99 360 0.0619 0.310
M 1 11.96 365 0.0883 0.185M 2 2.90 402 0.0602 0.519M 3 5.03 458 0.4400 2.0M 4 2.10 406 0.0402 0.479M 5 4.99 345 0.0312 0.156
Nickel
B 1 11.96 345 0.753 5.7
aNormalized togram of Li2 0.
1.51 cm2 area and wall thickness of 0.038 cm per
per year; the calculated rate is 4.5% per year. The copper measurementsfor tritium loss are scattered partly because of thermocouple and/or con-troller problems. The copper results extrapolate to less than 1% loss peryear at temperatures below 425 C, but they are one to two orders higher thanexpected.1 4 The experimental measurements, plotted in Fig. I-10, show thetemperature dependence of the percentage tritium loss extrapolated to oneyear. The fit of the data to a linear curve for the two capsules is excel-lent. There are several possible reasons for the poor agreement between thecalculated and measured permeation results. These include existence of amixed lithia-cupric oxide and lack of equilibrium between LiOH, Cu2O, Cu,H20(g), and H2 (g). In the latter case, about 7.1 x 10-8 mol Cu2O should havebeen formed by the end of Run T-5 if there were no excess hydrogen or morereducing impurities inside the capsule. The Cu20 could form about threemonolayers inside the capsule. The surface layers of thin Cu2O films arelikely to be more stable than bulk Cu20 and thus likely to produce higherhydrogen pressures. We plan to test this hypothesis by adding Cu2 0 to thecapsule in future tests of copper. In our post-test analysis we will look forsigns of strain on the capsules caused by high internal pressure. If a mixedlithia-cupric oxide exists, the oxygen pressure should be lower than that inequilibrium with Cu2 0 and copper, and the hydrogen pressure should be higher.The higher hydrogen pressure would increase the permeation rate. Calculationsindicate that a mixed oxide stable to the extent of only 5 kcal/mol wouldaccount for the observed increase in tritium loss.
In the next set of tests, we plan to measure the permeation of hydrogenand tritium from lithia reacted with tritiated water and tungsten with andwithout a nickel liner. The tungsten tests will establish the effectiveness
28
0.5 T496
U t
-0.5H..
V,)V')0
E
S-
a)Ua)
0~
00)
13-1.5
441 394 352 315 282
Temperature 'C
LegendT series
O M series... ...fit
.............\ .
0
14 15 16
10'/T, K~17 18
Fig. I-10. Tritium Loss by Permeation throughCopper Cladding as Function ofTemperature (rate extrapolated toone year of tritium breeding)
of tungsten as a tritium barrier. Tests will also be made with a silver-linednickel capsule. The oxygen pressure over silver is higher than that in equi-librium with copper and copper oxide. This higher oxygen pressure shouldsuppress the hydrogen (and tritium) pressure and thereby decrease the tritiumloss rate.
D. Fusion-Related Research
A critical element in the development of the fusion reactor is theblanket for breeding tritium fuel. We are conducting several studies with theobjective of determining the feasibility of using lithium-containing ceramicsas breeder material. We are also conducting design studies of methods for
improving fusion reactor performance and neutron dosimetry and damage analysisof fusion materials in neutron facilities.
1. Thermodynamics and Kinetics of Breeder Materials(A. K. Fischer)
Adsorption of H20(g), dissolution of OH~, and evolution of 1 20(g)
from LiAlO2-H20(g) are being measured to provide thermodynamic and kineticdata for these processes at high temperatures (400-600 C). Such data relateto the tritium retention and release in candidate ceramic tritium breedermaterials such as LiA102.
Adsorption measurements are made by the frontal analysis techniqueof gas chromatography, and solubility data are obtained by measurement of theuptake of H20(g) after the adsorption is finished. The technique wasdescribed in the previous report.'5
-1-
i i I i - i
.............
29
Until now, the analysis of the adsorption-dissolution curves hasfollowed the common assumption that the adsorption process is considerablymore rapid than the dissolution process. As a result, only a negligible con-tribution was expected from dissolution when compared to the apparent quantityof adsorbed H20(g). The observed sharpness of the breakthrough points hastended to support this view. Despite this, for high-temperature systems suchas those being studied, in which diffusion can become more rapid, it wasconsidered worth testing the assumption. One way would be to repeat the mea-surements with a sample of different surface area-to-volume ratio. However,in addition to such measurements being expensive, there would arise questionsof accurate reproduction of all other factors that might be influential; forexample, more material of different particle size from the same batch would bedesirable but is not available.
Another way of testing the assumption is to utilize the kineticinformation from the existing data to assess the diffusional contribution.A procedure to do this involves fitting an equation to the dissolution curveover the time interval immediately after adsorption and at least equal inlength to the adsorption time. The resulting expression for the dissolutionrate is then used to calculate the amount of dissolution occurring during theadsorption period. In effect, the procedure is to back-extrapolate the dis-solution to the time of sample injection. To apply this procedure, it isnecessary to supply a correction for the point-by-point readings of the wateranalyzer. These corrections are for the response curve of the water analyzerin the "step-up" mcde of measurement ("step-up" refers to a step change frompure helium to helium containing the particular level of H20(g) being used).The necessary data were gathered and analyzed by this procedure for the vari-ous flow rates used in the measurements. An initial check of the previouslyreported runs showed that the dissolution correction was indeed significant.Therefore, each one of the runs was reanalyzed according to this approach.The effect of the correction is to diminish the previously stated adsorptionfractions and to increase the previously stated solubilities.
From the raw data, before applying the correction in the mannerdescribed above, it appeared that the adsorption fraction at 5000C was greaterthan at 400C. Further experimental evidence was sought to resolve thispoint. Additional measurements of adsorption at 4000C and at 5000C were madeat similar partial pressures of H20(g) and closely spaced in time (within aday). The results reinforced the earlier conclusion about greater adsorptionat 500C than at 400 C. The new data and the earlier data were pooled. Therationale for the apparently inverted relationship between the 400 and 500*Cisotherms is given below.
Figures I-11 to -13 show the adsorption isotherms for 400, 500, and600 C, respectively, after the data points were corrected for dissolution.The curves report the fraction of the surface, 6, covered by 0H- as a functionof partial pressure of H20(g), pH O. The isotherms are of the Freundlichtype, as reported earlier. For clarity of the relationships of the isothermsto each other, they are plotted without data points in Fig. 1-14. The 4000Cand the 6000C isotherms are nearly parallel, with the respective regressionequations being:
logO = (0.593 0.141) + (0.517 0.18 2 )log(pH2o) (I-5)
30
-4.5 -4 -35
Log(p, H20, atm)
0-..
-0.2-]
-0.4 -]
-0.6 -j-0.8
-1.2-
-1.4-
-1.6 -
-1 8--2
-2.2-
-2.4
-2.6-
-2.8 -
-I-3
Fig. I-11.
-4.5 -4 -3.5
Log(p, H20, atm)
Fig. 1-12.
Adsorption400 0 (673
oi OH- on LiA102 atK)
Adsorption of OH- on LiA1Oa at5000C (773 K)
-5 -4.5 -4 -3.5
Log(p, H 20, atm)
o)0-J
-0.2-
-0.4-
-0.6-
-0.8-
-1-
-1.2 -
-1.4 -
-1.6-
-1.8 -
-2-
-2.2 -
-2.4 -
-2.6 -
-2 8-
-5
Fig. 1-13.
-4.5 -4 -3 5Log(p, H20, atm)
Fig. 1-14.
Adsorption of 0H- on LiA1Oa at6000C (873 K)
Adsorption4000C (6736000C (873
of OH- on LiA1O2 atK), 5000 (773 K),K)
0-J
Q)0-J
0O
C)0
-0.2-
-04 -
-0.6 -
-0.8-
-1 -
-1.2 -
-1.4 -
-1.6-
-1.8 -
-2-
-2 2 -
-2 4 -
-26--2 8 --28-
0
Temperature673K
773 K
873 K
-4. .
0-
-0.:-
-0.4-
-0.6 -
-0.8-
-1-
-1.8 -
-2 -
-2.2-
-2.4-
-2.6-
-2.8-
-1 1
n n
G J y
J
3
31
and
loge = (0.376 0.113) + (0.593 0.1 57)log(pH2Q) (1-6)
The regression equation for the 5000C isotherm is
logO = (0.055 0.018) + (0.359 O.02 4)log(pH2o) (1-7)
These curves show greater adsorption at 5000C than at 4000C. Theapparently anomalous observation of higher adsorption at a higher temperatureis a fairly common finding and was explained by Taylor in 1931.16 It had beenobserved that, for some systems, experimentally measured adsorption isobarsfollow the curve ABCD in Fig. 1-15. The explanation for this curve followsfrom the existence of two separate adsorption processes which, if they couldbe separated, would follow the isobars ABE and FCD. The isobar FCD reflects aprocess with a relatively high activation energy compared to that of processABE, which could even be unactivated. Starting from A, the system shows theexpected decline in the amount adsorbed as temperature rises. The temperatureis too low for process FCD to be appreciably activated. At B, the temperature
-) A
.0Fig. 1-15.
Schematic of Adsorption Isobarsfor Two Activated Processes oc
E
B
E D
Temperature
has risen enough so that sufficient activation energy is supplied to theprocess represented by the FCD isobar for that process to start contributingto the total amount adsorbed. A region of increasing adsorption with risingtemperature has been entered. This second process becomes increasinglydominant as adsorption rises with increasing temperature until the system isdescribed essentially by it alone. Then, adsorption again declines withincreasing temperature in the expected way. A relevant example of increasingadsorption with rising temperature is the report by Gruber'7 on the adsorptionof H2 on v-A1203 : the amount adsorbed increased by a factor of three from 300to 5000C.
32
In the present case, adsorption of H20(g) on LiA102 is viewed asinvolving two kinds of chemisorption. An atomic basis for such a distinctioncould be that one kind of chemisorption involves lithium ions and adjacentoxides, and the other kind involves aluminum ions and adjacent oxides. Slopesof 0.5 observed for the 400 and 6000C isotherms are supportive of dissociativechemisorption in which one molecule of H20 forms two adjacent OH- groups. Theobserved lower value of slope for the 5000C curve is expected if additiveeffects from an additional process are involved. Evidence consistent with twosurface processes in the evolution of water vapor from LiA102 was reportedearlier.i5 During heatup after a run, the water evolution curve showed adouble peak, which was interpreted as showing that different kinds of siteswere involved. Because more than one isotherm for a given process is neededif one is to calculate heats of adsorption, and the present situation providesonly one isotherm for each process, conclusions on heats of adsorption mustawait measurements of additional isotherms.
Corrected isotherms for hydroxide solubility in LiAl02 are presentedin Figs. 1-16, -17, and -18 for 400, 500, and 6000, respectively. In additionto the corrections originating in the dissolution occurring during adsorption,there is also a correction required for residual hydroxide that was discussedin the earlier report.' 5 This correction accounts for the hydroxide in equi-librium with the water content of the gas phase under baseline conditions.Applying this correction requires an assumption about the solubility ofLiAl5O8 in LiA102 . That is, it is assumed that, for the small amounts ofLiAlsO8 involved here, the LiAl5 O8 is a solute in LiAl02 rather than aseparate phase. With this assumption, the expected slope is 0.4 in a plot oflog(xoH-) vs. log(pH2ao). Therefore, corrections for residual 0H- were madefor each of the data points to give slopes of 0.4. (If LiAl508 were aseparate phase, a slope of 0.5 would be appropriate.) The regressionequations for OH~ solubility at 400, 500, and 6000C, respectively, are:
log(xoH-) = (-2.420 0.054) + (0.401 O.O57)log(pH2a) (1-8)
log(xoH-) = (-2.664 0.128) + (0.402 O.1O9)log(pH2 0) (1-9)
log(xoH-) = (-2.835 0.315) + (0.401 0. 5 2 4 )log(pH 2 o) (I-10)
The regression lines are plotted together in Fig. I-1y to indicate the solu-bility relationship at the three temperatures. Solubility is seen to declinewith increasing temperature. Preliminary findings from the measurements withH2 suggest that there might be a need to reinterpret the solubility data.Further comment is premature, except to caution that these solubilities arestill tentative. However, for matters as they stand, the heat of solution maybe estimated from these isotherms to be -5.5 kcal per mole OH-. Isobars forH20 partial pressures of 10~8, 10-5, and 10-4 atm are shown in Fig. 1-20. Thelinearity in these curves supports the corrections that were applied forresidual hydroxide.
-.3
33
-4 2
-4 4
-4.8
-5
S
"
-5 -4.5 -4 -3 5 -
Log(p, HO, atm)
Fig. 1-17.
Hydroxide Solubility in4000C (673 K)
LiA102 at Hydroxide Solubility in LiA1O2 at5000 (773 K)
-45 -4 -3.5
Log(p, H20, atm)-3
x0
C0Uco
L
0-J
-3 8
-4-
-4.2-
-4.4 -
-4.6 -
-4.8-
-5 -4.5 -4 -3.5
Log(p, H20. atm)
Fig. 1-18.
-3
Fig. 1-19.
Hydroxide Solubility in LiA1O2 at600 C (873 K)
Hydroxide Solubility in LiA1Oa at4000C (673 K), 5000C (773 K), and6000C (873 K)
-4- -4 -
x.- 2-
c
CO-44
0 -4.6-
0-.J
-4.8-
a)
COU
a)0
0-J
-4.5 -4 -35
Log(p, H 20, atm)
"
Fig. 1-16.
-4 -
X
C0
Q
C)C
0-J
-42-
-46-
-4 8 -
x
Temperature
673K773K
873K
-5-5
-3.8 _-3.8
-- ti.
5 .
-4 4
34
-4-
-425 -
-4.5
sIFig. I-20.
-4.7 Hydroxide Solubility in LiA102at Three H20 Partial Pressures
0E -5
H 20 Partial Pressure-5.25- o E-e aim
C IE-5 atm
0 1E-4atm
1.1 1.15 1.2 1.25 1.3 1.35 1.4 1.45 1.5
Reciprocal Temperature, 1000/K
2. Modeling of Tritium Behavior in Ceramic Breeder Materials(J. P. Kopasz)
Temperature has a large effect on tritium release from ceramicbreeder materials.'8 The temperature dependence of the tritium release mayprovide some insight regarding which mechanisms are rate controlling underdifferent conditions. A second variable which also has a large effect ontritium release is the sample grain radius. The dependence of the tritiumrelease predicted from our diffusion-desorption model was investigated forLi2 0 using the diffusional parameter (D) reported b Guggi'9 and the desorp-tion rate constant (K) reported by Kudo and Okuno For samples with a smallgrain radius, the diffusion time is short and tritium release moves towardbeing desorption controlled. For large grains, the diffusion time is long andthe release moves Loward being diffusion controlled.
To determine where tritium release is diffusion controlled, the dif-ference between the tritium release predicted from the pure diffusion modeland the diffusion-desorption model was plotted as a function of temperaturefor two grain radii (Fig. 1-21) and as a function of grain radius for twotemperatures (Fig. 1-22).
At 4000C (Fig. 1-22) the diffusion model overestimates the tritiumrelease at short times. As the tritium release approaches steady state, thisoverestimation decreases and eventually becomes zero. For large grainsamples, the diffusion model initially gives a better approximation of thetritium release than for small grain samples. As time proceeds, the diffusionmodel becomes a better approximation for small grain samples than for thelarge grain samples. This is more readily seen in the plot at 500*C.
For the 10 pm grains (Fig. 1-21, bottom) at low temperatures(400-475 C), the diffusion model is a poor approximation to the diffusion-desorption release. The overestimation in tritium release predicted by thediffusion model increases with increasing time, reaches a maximum, then
35
0.3 50 pm
0.7
0.5
0.4
0.Z
O 0.1
02
OS
4 Z.1 0
400
4 . 10 m
W0.'0.
5 -J
Q .S
ti 0.4
fr 0.1
03
l gl
43 q
400o
Fig. 121. Difference between Fractional ReleaseCa. ulated Using a Diffusion Model and aDiffusion-Desorption Model as a Functionof Time and Temperature for 10- and 50-pmGrain Sizes
36
W
0.9Ui 400 OC
W 0.8UJ
N 0.7
0.6
U_
U
0.5
0.4
000.3
0.2
0.1
O 0
'4g
0.0
RADIU Cri) 0.003 30.002s
0 0 0 1 '0 0
0.000
W
.9 500 c
d0.8
E-0.7
0.6
0.5
1 0.4
Q 0.3
0.2
0.1
O 0
00.000
Fig. I-22. Difference between Fractional ReleaseCalculated Using a Diffusion Model and aDiffusion-Desorption Model as a Functionof Time and Grain Radius for a Sample at400 and 50000
37
decreases as the steady-state release is approached. Desorption is thedominant mechanism controlling tritium release in the low temperature region.At high temperatures, the diffusion model gives a good approximation of thediffusion-desorption release, except for very short times.
For 50 pm grains (Fig. 1-21 top) the diffusion distance is longer,and as one would expect, the diffusion model is a better approximation to thediffusion-desorption model. At low temperatures, the difference between thetwo models increases with increasing time, then reaches a maximum. Thismaximum is reached more rapidly at high temperatures, then decreases to zerowith increasing time. At temperatures above 525C diffusion is the dominanttritium release mechanism.
When attempting to use the diffusion-desorption model or the purediffusion model, one encounters difficulties because of the large variationsin the tritium diffusivities for the lithium ceramics which appear in theliterature and the lack of data for the desorption rate constants. Lithiumoxide is probably the material for which the tritium diffusion data are thebest, yet there is still some disagreement over the diffusion coefficients.Kudo and Okuno report diffusion activation energies of 154 kJ/mol in Li2 0pellets2o and crystals.21 Activation energies for single crystals have beenreported as 101 kJ/mol by Tanifuji et al.2 and 81.7 kJ/mol by Guggi et al.19The work by Guggi and Tanifuji appears to be the most carefully done, and webelieve the diffusivity of tritium in Li2 0 lies between these two values. Anadditional factor pointing to these values is the lithium self-diffusion data.Tritium is expected to diffuse by the same or similar mechanism as lithium inlithium oxide. This suggests that the activation energy for diffusion shouldbe similar for the two materials. Lithium self-diffusion studies by Oei andRichterung23 and Oishi et al.24 report activation energies for lithium self-diffusion of 107 and 98 kJ/mol, close to the activation energies for tritiumdiffusion found by Guggi and Tanifuji. The desorption activation energies forthe desorption of tritium from Li2 0 reported in the literature vary from 120to 150 kJ/mol.2 0'2 5 However, because a desorption activation energy of71.2 kJ/mol was recently reported,2 8 there is still considerable doubt as tothe magnitude of the activation energy of desorption of tritium from Li2 0. Itis possible that the desorption activation energy is dependent on the amountof surface coverage on the ceramic material. This may be the cause of thevariations in the reported values and suggests that experiments designed toobtain the desorption rate constants are needed and that reliable desorptionrate constants cannot be obtained from in-pile purge flow experiments.
For the solid lithium breeder materials other than Li2 0, there areeven fewer data for the diffusion and desorption parameters. For LiA102 thereis a wide variation in the reported data, and few data for the silicates canbe found. As a guide to choosing diffusion activation energies for thesematerials, we have used some studies of lithium self-diffusion. The activa-tion energy for lithium self-diffusion in Li2SiO3 is very close to that foundfor LiAl02,
2 7 suggesting that the activation energy for tritium in the twomaterials should be similar. In addition, these activation energies wereabout twice that observed for lithium self-diffusion in Li2O.2 7 Using theactivation energy for tritium diffusion in LiA102 reported by Clemmer et al.(150 kJ/mol),'8 we calculated the tritium release profiles for the LISA-1experiment with pure helium purge gas.2 8 The diffusion-desorption model was
38
used for this purpose. Good fits of the predicted and measured values wereobserved for the Li2 SiO3 samples (Fig. 1-23) and only fair fits were obtainedfor the LiA102 samples. It appears that some mechanism other than diffusion-desorption may be involved in tritium release in the aluminate sample.
Data for tritium release for five runs from the CRITIC experimentwere sent to us by Dr. Verrall at Chalk River. The CRITIC experiment is anin-pile tritium release experiment using lithium oxide with an average grainsize of 30 pm. The data from the release measurements are unusual in that,when the temperature was increased, the tritium release initially dropped (seeFig. 1-24) in all runs but one (run 2). This run (Fig. 1-25) is the one whichattained the highest temperature (680 C). Maximum temperatures for the otherfour runs were 549 0C (run 1), 572 C (run 3), 537C (run 4), and 616C (run 5).Attempts to fit the release curves for the temperature increases to a simplediffusion-desorption model failed for all but one run (run 2), the one whichdid not show the decrease in tritium release. Using the DISPL computercode,2 9 we were unable to fit these release profiles for diffusion-desorptionmodels with the temperature of the sample rising nonuniformly (i.e., tempera-ture of the surface rose instantaneously and the temperature of the bulk roseslowly, and vice versa). A model using a positive energy of desorption(i.e., negative activation energy) provided a release curve with a similarshape to that obtained in the CRITIC experiment (see Fig. 1-26). A positiveenergy of desorption here is not physically realistic. However, if there aretwo mechanisms or sites for desorption, each with different activationenergies, and if the desorption changes from one process to the other, it mayappear as if the energy of desorption were positive. This is a probable ex-planation for the observed release curves. Several examples of adsorption and
3.5-
Calculted3-
Observed
2.5
Fig. 1-23.
Predicted and Measured Tritium E 2-
Release for Li2SiO3 in LISA 1,
Experiment 5-.5-
IdI
'//
0.5
0 200000 400000 600000 800000
TIME, s
39
50-
45-
40-
35-
30-
25
20000 40000 60000
TIME, s
70
60
50
40
30 -
20 -
10 -
010 12000 24000 36000 48000
TIME, s
Fig. 1-25. Tritium Release Obtained fromCRITIC Experiment Run 2
I
E
LU
Ul)
F--
2 00
(-)W
L~J-J
E
60000 72000 64000
1 1
Fig. 1-24.
Observed Tritium Release for aTemperature Increase from 455 to5490C
.v
40
50-
E 45-
U]L1J
Calculated Tritium Release in 40CRITIC Experiment Using Positive ccEnergy of Desorption
c: 35
300 50000 100000 150000
TIME,s
desorption processes have been found for which there appears to be two dif-ferent adsorption or desorption sites. Hydrogen adsorption on alumina showsadsorption curves with an energy of adsorption at 150-399C of 7-12 kcal/moland an energy of adsorption above 300C of 12-18 kcal/mol.3o Recent work onwater adsorption by LiA102 suggests that there are two types of sites for H2 0adsorption on LiA102 . We will employ a model with an activation energy ofdesorption that is dependent on surface coverage to determine if such a modelcan reproduce the CRITIC results.
The effects of adsorption and diffusion in the gas phase were addedto the model. A model for adsorption was developed based on work in theliterature for H2 adsorption on A1203 , since no data could be found for H2adsorption on the materials of interest (Li2O, Li2 SiO3 , and LiA102). Adsorp-tion isobars have been reported indicating that two types of chemisorptionoccur.3 1 From these data, the following equations were derived:
Vad = 3.053 x 10~3 exp(3.153 kcal/RT) for T <3520C (I-11)
Vad = 29.8 exp(-8.279 kcal/RT) for T >352C
where Vad is the volume adsorbed (cm3 /g). For a hydrogen pressure of200 mm Hg at 4500C, the calculated hydrogen surface concentration is3.65 x 10-12 mol/cm2 or 4.54 X2 /H atom. Using a surface thickness of 10 X,this converts to a concentration of 3.65 x 10-5 mol/cm3 . Other reports ofhydrogen adsorption on alumina report monolayer coverage would correspond to6.6 X /H atom. 2 The hydrogen surface coverage in the model increased as thefourth root of the hydrogen pressure based on the observation of thisdependence for hydrogen on alumina.3 3
41
The tritium diffusing in the gas phase was treated as HT. The HT isalso adsorbed from the gas phase onto the surface with an adsorption rateequal to that for an equivalent partial pressure of H2. The model considerstritium being generated in a cylindrical sphere, with diffusion occurringradially and adsorption and desorption occurring along the whole length of thesphere. Attempts to run this model on the VAX computer in the CMT Divisionhave been unsuccessful due to the coarse computational mesh which had to beemployed in the differential equation solver code for the two-dimensionalcase. We are now altering the code to allow for a finer mesh and transferringthe code to a more-powerful computer to handle the larger number ofcomputations needed with a finer mesh.
A related effort involved calculating the effects of impurities inlithium oxide on the tritium diffusivity. A model presented by Tam et al.3 4
was reviewed and altered slightly to take into account some conflicting dataon the lithium diffusivity in lithium oxide and to concentrate on the effectsof the impurities on diffusion. The new model assumes that desorption is muchmore rapid than diffusion, resulting in diffusion-controlled release. Thisallows us to separate the effects of impurities on diffusion from theireffects on desorption and simplifies the mathematical model, thereby loweringthe computation time.
The model is based on the assumption that tritium diffuses as alithium-vacancy/tritium complex in Li2a0. There are three species to consider:the lithium vacancy, the tritium, and the tritium/vacancy complex. Thetritium and the lithium vacancies are generated at an equal rate in the solid.In addition, initially, there are vacancies present due to impurities and dueto thermal motion in the solid. The number of vacancies due to thermal motion(n9) causing the formation of a Schottky defect can be calculated by thefollowing expression:
n, = N exp(-AH,/2RT) (1-12)
where N = number of lithium atoms. The energy of formation of the defect,AH., can be determined from the lithium self-diffusion data if the diffusivityis determined in the intrinsic region, where the vacancies due to thermalmotion overwhelm those due to impurities, and in the extrinsic region, wherethe vacancies due to impurities overwhelm those due to thermal motion. In theintrinsic region, a random walk model of diffusion gives the diffusivity (D)as
D = g a2 v [N exp(-AH,/2RT) exp(-Hm/RT)] (1-13)
where g = geometric factora = jump distancev = jump frequency
Hm = energy of migration of the defect
When impurities are present, the diffusivity can be estimated by
D = g a2 [nd + N exp(-AH,/2RT)] exp(-Hm/RT) (1-14)
where nd is the number of vacancies present due to impurities. Measurement ofthe diffusivity in the extrinsic region gives Hm, while that in the intrinsicregion yields AH,.
42
The two groups of data on lithium self-diffusion in Li2O from theliterature are in disagreement. Oei and Richterung have reported an intrinsicactivation energy of 1.1 eV (106 kJ/mol) and an extrinsic activation energy of0.55 eV.23 Their work is supported by other reports of activation energies inthe extrinsic region varying between 0.3 and 0.75 eV, depending on the impuri-ties present and their amounts.2 7 ,3 5 However, an earlier work by Oishi et al.arrived at an intrinsic activation energy of 243 kJ/mol and an extrinsicactivation energy of 97.9 kJ/mol.2 4 Using the Oei and Richterung data, onearrives at an Hm of 53 kJ/mol and AH of 106 kJ/mol. Alternatively, the Oishidata result in an HM of 97.9 kJ/mol and AH of 291 kJ/mol. This implies thattritium diffusion is in either the intrinsic or the extrinsic region, depend-ing on which data are correct. At present, calculations are being performedusing both values of H and AH,.
Some initial results from calculations using the Oei and Richterungdata have been obtained. These calculations have not yet reached the point ofsteady-state tritium release; however, a comparison of the inventories atidentical times gives a good indication of how the inventories will compare atsteady state. The inventory is unaffected at 400C when impurities are addedto the Li2O up to the level of 100 ppm, which is an order of magnitude greaterthan the number of vacancies present due to thermal equilibrium.
3. Design Studies of International Thermonuclear Experimental Reactor(P. A. Finn)
The international thermonuclear experimental reactor (ITER) is a lowpower (<900 MW) reactor whose main purpose is to provide a fusion environmentfor engineering tests of fusion reactor systems.
For ITER, we initiated an assessment of radiolysis that would beproduced in a 2-10 M aqueous lithium nitrate solution, a blanket designconcept under consideration. We are assessing the effect of radiolysis on thecomposition of a concentrated lithium salt solution by using information inthe literature from similar systems. Since comprehensive radiolysis theorydoes not exist for concentrated salt solutions at high linear energy transfer,we are making worst-case estimates of radiolysis effects for dilute saltsolutions at low linear energy transfer. Staff at Atomic Energy of Canada,Ltd. (AECL) have estimated that as much as ~300 L/s of hydrogen could beevolved.
4. Tritium Control Experiment(P. A. Finn and E. H. Van Deventer)
The management of tritium transport within high-temperature fusionblanket systems is required to achieve an environmentally acceptable fusionfacility. A management method advocated for controlling permeation losses isthe oxidation of the tritium to tritiated water. If oxidation of the tritiumoccurs, the molecular tritium partial pressure will be reduced. For fusionblanket systems, this oxidation must be >99% complete for effective managementof tritium transport. The purpose of this experiment is to study the heter-ogeneous reaction of oxygen on the outside of a stainless steel tube (doublewalled) with tritium permeating from the inside. A tritium/helium gas mix-ture is circulated through the inner tube, and an oxygen/helium mixture iscirculated through the outer tube. The amount of tritiated water formed was
measured for a range of oxygen concentrations (<1 to 2000 ppm) in the helium
43
gas and temperatures of 350-550 C. The goal of this experiment is todetermine the effect on water yield of tritium concentration, oxygen con-centration, and temperature. The experimental apparatus was described in thepreceding report in this series. The oxygen concentrations and temperatures,along with the yields of molecular tritium and tritiated water, are given inTable 1-13.
To test the effect of tritium content in the tritium/helium gasmixture, we completed a series of runs at 4000C in the range 10-3 to 1300 ppmoxygen at ~5 x 10-4 pCi/min tritium, a rate much lower than that in otherruns. For the 4000C runs, the HTO yield with 1300 ppm oxygen was 84.8%, whichis significantly lower than the 97.8% yield found for the conditions 550*C,1200 ppm oxygen, and 7 x 10-2 Ci/min tritium. A yield of 94.5% was found forruns at 350 C, 1200 ppm oxygen, 3 x 10-3 ,Ci/min tritium. These resultssuggest that the lower yield at 400C is not due to temperature effects but tothe lower tritium rate of '5 x 10-4 pCi/min. Because the amount of tritiumcollected in the 4000C run was low, we plan to repeat runs at 4000C with atritium content comparable to that in the 5500C run. This should verifywhether temperature is important at 4000C.
To test the effect of temperature, we completed runs at 3500C and5000C. The HTO/HT ratios obtained in these runs are given in Table 1-13 alongwith results for runs at 400 and 550*C. We observed that '20 ppm oxygenappears to be required at all temperatures to achieve >80% conversion of thetritium to tritiated water.
Figure 1-27 shows the percent yield of tritiated water versus oxygenconcentration for the runs in Table 1-14. The run at 12 ppm oxygen and 550Cappears as a sharp break in the line for the 5500C data. The two runs at500 C appear to bracket a similar break. We plan to repeat runs at 5500C toconfirm that this break occurs since the residence time for the 12 ppm oxygenrun was 20-25% times shorter than that of the other 5500C runs. The curve forthe 4000C data is below that for the runs done with higher tritium con-centrations. We cannot yet explain the significance of this observation.
The tritiated water yields were also plotted versus time assuminghalf-order dependence for water formation from the total tritium content.The slopes and intercepts derived from a least-squares fit are shown inTable 1-15. The fact that the slopes are not constant may indicate that theslope depends on the oxygen level or on the type of metal oxide present at agiven oxygen level. In cycling between low and high oxygen concentrations,the metal oxide at the surface could be different, thus giving rise to part ofthe differences noted in the slopes. The order in which the different oxygenconcentrations occurred resulted in alternate oxidation and reduction of themetal oxide.
We began destructive analysis of test cells 1 and 2, in which ini-tial runs in this series were done, and an unused (blank) test cell. Theportions of the test cells sectioned for examination were the Type 316stainless steel thin-wall test section, which was exposed to the test tem-peratures, and the Type 304 stainless steel thick-wall tube weld to the thin-wall test section. Maximum changes in the metal and/or metal oxide wereexpected in these areas.
44
Table I--13. Ratios between Tritiated Water and Molecular Tritiumaas a Function of Oxygen Concentration at 350-550 C
TritiumCollected, pCi
Oxygen, Temp., Time, Total T, Ratio HTO Yield,
ppm C T2 0 T2 min pCi/min HTO/HT %20 350 19.8 2.9 8488 0.003 6.8 87.2
1200 350 17.0 1.0 5435 0.003 17.1 94.5
570 400 12.0 2.5 9934 0.0015 4.8 82.8<0.1 400 4.0 2.3 8183 0.0008 1.75 63.61300 400 3.4 0.6 7208 0.00056 5.6 84.8
30 400 3.4 0.8 9672 0.00044 4.,2 80.7
20 500 214.7 33.1 3056 0.081 6.5 86.74 500 629.2 188.2 8247 0.099 3.3 77.0
100 550 31 4 397 0.09 7.8 88.6700 550 320 14 1327 0.25 22.8 95.8400 550 718 66 6398 0.13 10.9 91.640 550 665 97 7225 0.11 6.9 87.3
2000 550 529 11 7309 0.08 48.7 98.00.001 550 435 398 8533 0.10 1.1 52.41200 550 390 9 6798 0.07 43.6 97.8
0.001 550 668 617 18713 0.08 1.1 52.420 550 442 65 10066 0.05 6.8 87.212 550 130 88 4067 0.05 1.5 60.0
aRuns listed in order done.
105
95
e5
75
65
55
450.0001 0.001 0.01 0.1 1 10
Oxygen Concentration, ppr
Fig. 1-27. Tritiated Water YieldOxygen Level for Runs
100 1000 10000 100000m
as Function ofat 350-550 C
v
10*0d-i
x
Temperatureo = 350 CA = 400 C - Low Tritium -+ = 500 Cx = 550 C
45
Table 1-14. Summary of HTO/HT Ratiosa at Different Temperatures
HTO/HT Ratios
Oxygen,ppm 5500 5000 400*0 3500C
0.001 1.1 - -0.1 - - 1.75 -
4 - 3.3 - -12 1.5 - -20 6.8 6.5 - 6.830 - - 4.2 -40 6.9 - -
100 7.8 - -400 10.9 - -570 - - 4.8 -700 22.8 - -1200 43.6 - - 17.11300 - - 5.6 -2000 48.7 - -
aA ratio of 22.8 equals 95.8% conversion, whereas a ratio of 6.8equals 87.2% conversion.
Table 1-15. Slopes and Intercepts for Curves of Tritiated Water Yield vs.Time (half-order dependence assumed)
Oxygen Temp,a Slopeb Intercept Correl. Wtd. STDCppm 00 Coef.
20 350 6.4(-4) -4.32 0.992 0.151200 350 3.8(-4) -5.09 0.966 0.15570 400 2.2(-4) -3.79 0.998 0.033<0.1 400 1.2(-4) -2.53 0.997 0.0191300 400 1.7(-4) -1.99 0.994 0.026
30 400 1.1(-4) -1.97 0.969 0.05620 500 3.1(-3) -16.2 0.969 0.624 500 1.9(-3) -29.2 0.993 0.47
100 550 9.5(-3) -6.3 0.94 0.38700 550 1.1(-2) -18.7 0.97 0.99400 550 2.7(-3) -28.5 0.99 0.6140 550 2.4(-3) -28.0 0.99 0.69
2000 550 2.5(-3) -24.2 0.97 1.040.001 550 1.0(-3) -28.9 0.996 0.731200 550 3.4(-4) -82.3 0.98 0.12
0.001 550 5.9(-4) -35.7 0.99 0.2620 550 1.3(-3) -22.4 0.99 0.5612 550 1.3(-3) -14.8 0.996 0.12
aExperimental conditions given for runs in order done.
bNumbers in parentheses indicate power of ten.
cWeighted standard deviation.
46
Post-test examination of the blank cell sample revealed that theinner wall of the 304 SS tube had a very rough texture with deep pits ( 50 mdeep), and the 316 SS had scattered inclusions or healed cracks between theinner and outer surfaces. In addition, the 316 SS sample was composed of 96pm grains in the bulk material and 16 im grains at the edges. No otherabnormalities were noted. The appearance of samples from test cells 1 and 2was identical to that of the blank cell.
Scanning electron microscopy (SEM) was used to examine the outersurface of an unpolished 316 SS sample from test cell 1. There was noindication of excess iron, i.e., preferential formation of an iron oxide, atthe surface. The elemental composition at the oxide surface (averaged forfive separate locations) was similar to that of the 316 SS. It appears thatfor test cells 1 and 2, which had limited use, the oxide thickness was toothin to be detected by SEM.
5. Dosimetry and Damage Analysis(L. R. Greenwood, C. A. Seils, and A. Intasorn)
a. Neutron Measurements at the IPNS Facility
Five experiments were conducted at the Intense Pulsed NeutronSource (IPNS) facility at Argonne to measure the neutron flux spectra. Oneexperiment was performed for J. Carpenter (IPNS) and B. Williams (Iowa StateUniversity) to determine the fast neutron spectrum in the EVS spectrometer.Previous measurements were made in July 1987 for periods of 4 and 59 h. Thickfoils were placed in the beam line at 5.7 m from the neutron source. Bare andgadolinium-covered foils were used to measure the neutron energy spectrum inthe thermal/epithermal ranges. However, these short runs were not sufficientto determine the weak fast neutron fluxes. Consequently, a longer irradiationwas conducted for 17 days during September 11-28, 1987. The foils measured'2.5-cm square and consisted of Fe, Au, Cu, Al, Co, Ti, and Ni withthicknesses between 1 and 35 mils (25 and 890 m). Following the irradiation,gamma spectroscopy was used to measure 22 different radionuclides. Theresults were then corrected for geometry, neutron and gamma self-absorption,and decay during irradiation. The results are listed in Table 1-16 along withthe earlier measurements, all normalized to the activation rate per incidentproton.
The activities were used to adjust the neutron flux spectrumwith the STAY'SL computer code (a generalized least squares adjustment code).Since there are no calculations of the neutron spectrum at this position, wetried various spectral shapes based on previous measurements within themoderator assembly. As expected, the spectrum is much softer in the presentcase, and we have assumed a smooth extrapolation to the very highest neutronenergies (above 40 MeV). The adjusted fluxes are listed in Table 1-17 andshown in Fig. 1-28. As can be seen, these measurements determine the fluxesto within 5-15% below 40 MeV. At higher energies, we were only able to mea-sure a few reactions, as shown by the energy sensitivity ranges in Table 1-16.The uncertainties thus increase to 24% in the 40-100 MeV range and exceed 100%at 100-450 MeV. In any case, these fluxes are quite weak, with only about
47
Table 1-16. Activity Measurements in the EVS Spectrometer at IPNS
Rate,atoms/atom- 90%
Reaction proton Uncert., % Energy Range, MeV
55Mn(n,7)58Mn 1.80 x 10-31 2.5 10-9-2.8 x 10-4+ Gd cover 7.03 x 10-32 2.4 5 x 10-8-5.8 x 10-4
197Au(n,7)198Au (thick) 2.91 x 10-30 2.4 10-9-6.9 x 10-5+ Gd cover (dilute) 2.94 x 10-30 2.1 2.6 x 10-7-3.0 x 10-5
llbIn(n,7)11emIn 3.26 x 10-30 2.3 10-9-2.8 x 10-8
63Cu(n,7)64 0u 7.07 x 10-32 2.4 10-9-2.0 x 10-3
+ Gd cover 1.88 x 10-32 2.6 5 x 10-8-5.5 x 1-358Fe(n,7)5 9Fe 2.07 x 10-32 2.6 10-9-2.8 x 10-459Co(n,a)60Co 5.86 x 10-31 2.0 10~9-1.0 x 10-454Fe(n,p)54Mn 3.52 x 10~34 5.5 2-2827A1(n,a)24Na 7.40 x 10-38 4.1 7-20
11bIn(n,n')ll1mIn 1.17 x 10-33 5.0 0.5-558Ni(n,p)58Co 3.85 x 10-34 2.0 1-1058Ni(n,2n)57Co 2.34 x 10-38 5.5 14-2846Ti(n,p)48Sc 5.58 x 10~35 8.0 3-3647Ti(n,p)47Sc 1.21 x 10-34 3.0 1-3648Ti(n,p)48Sc 3.18 x 10~36 9.0 6-24
93 Nb(n,2n)92mNb 1.89 x 10-36 3.4 10-2059 Co(n,2n)58 Co 3.39 x 10-35 8.0 12-28
'97Au(n,2n)1'9 Au i.05 x 10~34 4.0 9-20
Cu(n,*)58Co 5.13 x 10-37 4.0 24-120
Cu(n,x)57Co 1.70 x 10-37 11.0 32-240
Table 1-17. Flux Measurementsat IPNS
in the EVS Spectrometer
Flux,Energy Range n/m2'p Uncert., %
Thermal (<0.5 eV) 1.40 x 10-12 4.8
0.5 eV-0.1 MeV 4.14 x 10-12 8.9
0.1-1 MeV 8.56 x 10-13 10.7
1-5 MeV 3.47 x 10-13 8.3
5-10 MeV 1.84 x 10-15 7.8
10-20 MeV 4.37 x 10-15 6.5
20-40 MeV 1.83 x 10-15 14
40-100 MeV 2.38 x 10-16 24
>100 MeV 9.12 x 10-17 100
48
IPNS F3 5.7m
10
1CIw
z 102C
J 10LL
10 10' 10' 10 610 10 ' 10 2 10' 0 10' 102 10NEUTRON ENERGY,MeV
Fig. 1-28. Neutron Energy Spectrum Measured at 5.7 m inthe EVS Spectrometer on the F3 Beam Line atIPNS. Flux per unit lethargy (energy timesflux) is shown on y-axis. The initial spec-trum is shown as a solid line; the dottedline is the final adjusted flux.
57 n/cm2 's above 100 MeV at a beam current of 10 #A. Hence, the measurementsshould be adequate for determining fast neutron effects in the EVS Spectro-meter at IPNS.
b. Dosimetry and Damage Analysis for the Omega West Reactor
Neutron dosimetry measurements and damage calculations werecompleted for three experiments in the Omega West Reactor (OWR) at Los AlamosNational Laboratory. These irradiations were conducted by Hanford EngineeringDevelopment Laboratory (HEDL) and were designed to compare radiation damageproduced in a fission reactor spectrum with damage produced by 14 MeV neutronsat the Rotating Target Neutron Source (RTNS) II at Lawrence Livermore NationalLaboratory.
Three experiments were conducted by Howard Heinisch (HEDL) atOWR between May 21 and July 16, 1987. The experimental conditions weresimilar to those reported previously.3 ~-3 8 The irradiation histories andexposure parameters are listed below.
Experiment Dates Exposure, FPH
9 5/21/87-5/22/87 12.9710 5/27/87-6/5/87 58.1211 6/22/87-7/16/87 127.50
The exposures are listed in full power hours (FPH) at 8 MWreactor power. Small dosimetry capsules were included with each experiment.Each capsule contained wires of 0.1% Co-Al, Ni, Fe Ti, and 80% Mn-Cu. Each
49
wire was gamma counted at Argonne, and activation rates were computed consid-ering the irradiation histories and gamma absorption. The results are listedin Table 1-18.
Table 1-18. Activation Rates Measured in the OmegaWest Reactor (at 8 MW, 2%)
Activities, atom/atom-sec
Reaction Run 9 Run 10 Run 11
5SFe(n,7)59 Fe (x10-11) 6.44 6.35 5.98
s9Co(n,7)8 0Co (x10~9) 2.16 2.05 2.08
54Fe(n,p)54Mn (x10-la) 2.74 2.81 2.75
58Ni(n,p)58 Co (x10-i2) 3.61 3.63 3.4046Ti(n,p)48Sc (x10-13) 3.73 3.77 3.60
s5Mn(n,2n)54Mn (x10-i5 ) 8.42 8.46 8.27
Neutron energy spectral adjustments were then made for eachirradiation using the STAY'SL computer code starting with a previously mea-sured spectrum.3 The resultant fluxes and fluences are listed in Tables 1-19and 20. As can be seen in the tables, the activities and fluxes for all threeruns are in agreement within a few percent. These results also agree with theprevious run numbers 1-4, also with helium cooling.
Damage calculations were performed for each run using theSPECTER computer code39 and the results are listed in Table 1-21.
Table 1-19. Neutron Fluxes for Omega WestReactor (at 8 MW, 10%)
Flux, 1013n/(cm2 's)
Energy Run 9 Run 10 Run 11
Thermal (<.5eV) 7.28 6.94 6.92
0.5eV-0.1 MeV 4.74 4.58 4.80
>0.1 MeV 5.45 5.49 5.43
>1 MeV 2.68 2.72 2.67
Total, MeV 17.47 17.01 17.15
50
Table 1-20. Neutron Fluences for Omega WestReactor ( 10%)
Fluence, 1018 n/cm2
Energy Run 9 Run 10 Run 11
Thermal (<.5eV) 3.40 14.50 31.75
0.5 eV-0.1 MeV 2.22 9.57 22.04
>0.1 MeV 2.55 11.47 24.92
>1 MeV 1.25 5.69 12.24
Total, MeV 8.16 35.55 78.72
Table 1-21. Damage Parameters for Omega West Reactor
Run 9 Run 10' Run 11a
Element dpa He dpa He dpa He
Al 3.27 1.07 14.72 4.83 31.98 10.49
Ti 1.97 0.99 8.88 4.45 19.29 9.67
V 2.23 0.04 10.03 0.17 21.80 0.36
Cr 2.01 0.28 9.04 1.27 19.65 2.75
Fe 1.78 0.49 8.00 2.21 17.38 4.80
Ni 1.89 7.57 8.50 34.04 18.47 73.96
Cu 1.72 0.41 7.76 1.85 16.85 4.01
Nb 1.74 0.09 7.82 0.42 16.98 0.91
aUnits are dpa x 10-3 and He in appb.
Additional experiments are in progressexperiments are planned with water cooling to confirmbetween water and helium.
in OWR. In particular,the spectral differences
c. Neutron Yield and Spectral Measurements at IPNS
Foil activation experiments have been conducted to measureangle-dependent neutron yields and energy spectra from range-thick Al and 2 3 8Utargets bombarded with 50, 113, and 256 MeV protons and 98 MeV deuterons. Theproton irradiations were conducted at IPNS at Argonne. The 50 MeV irradia-tions were performed on December 18, 1987, in the Neutral Particle BeamFacility at the IPNS. The aluminum experiments at 113 and 256 MeV were per-formed on December 10, 1987. The 113 MeV proton irradiation of depleted
51
uranium occurred on January 27, 1988, and the 256 MeY run was on February 24,1988. The higher energy irradiations required the use of the Rapid CyclingSynchrotron and were performed in the Beam Dump area in the main IPNS beantransport tunnel.
Deuteron irradiations were performed at the Michigan StateUniversity Cyclotron in East Lansing, Michigan, on February 20-22, 1988. Inthis case, a 98 MeV deutron beam was stopped in thick Al and 2 3 8U targets.These irradiations were performed in collaboration with the Air Force WeaponsLaboratory to compare our foil activation results with their time-of-flightneutron spectroscopy.
In all cases, the purpose of these experiments is to measurethe neutron yield and energy spectra at various angles using the multiple foilactivation technique. In particular, we want to demonstrate the feasibilityof our techniques at higher neutron energies (up to 250 MeV) using multiple(n, xn) and spallation reaction products to measure higher-energy neutronyields. These results will then be compared with neutronic calculations andtime-of-flight measurements to assess the reliability of the foil activationtechnique. Based on these comparisons, we then plan to submit proposals forfurther foil measurements as a quicker, cheaper alternative to time-of-flightexperiments on a variety of target materials over a large range of proton anddeuteron energies.
In each case, stacks of foils (~2.5 cm OD, 25-1000 m thick)were irradiated at distances between 15 and 35 cm from the center of thetarget at angles between 0 and 150 degrees from the incident beam. Hundredsof foils were then gamma counted to determine activation rates for 20-30different radionuclides at each position. These activities will then be usedto adjust the neutron energy spectra using the STAY'SL computer code withhigh-energy activation cross reaction which we have measured previously at theIPNS.
All of the irradiations at IPNS had beam currents between 1 and5 A for times of 1 or 2 h. Foil packages were located at angles between 0and 135 degrees at distances of 12 to 30 cm. Larger distances were needed atthe higher energies due to the large sizes of the range-thick targets. The50 MeV targets measured ~5 cm square while the higher energy targets werecylinders of ~10 cm diameter. The foil packages measured ~2.5 cm in diameterand varied between 0.30 and 0.63 cm in thickness. Each package containedfoils of Al Au, Cu, Co, Nb, Ni, In, Fe, Ti, and Zr; foil thicknesses variedbetween 25 and 890 m. Each foil packet was wrapped in thin aluminum foil andsupported on an aluminum rod.
Following each irradiation, the foils were returned to CMT forgamma analysis. Several reaction products were measured from each foil, andit was often necessary to recount samples at different decay times to measureboth short- and long-lived decay products. The Gamanal computer code was thenused to analyze each spectrum and to determine disintegration rates for eachradionuclide. These activities are now being corrected for geometry, self-absorption, and decay during irradiation. The corrected activation rates willthen be used to determine the neutron flux spectra at each energy and angleusing STAY'SL.
52
One of the safety concerns for these irradiations wastemperature control. This problem is complicated by the need to minimizesupport structures around the target, since extraneous materials may perturbthe neutron spectra. Water cooling would be quite sufficient to dissipate theheat; however, water would severely moderate the neutron spectrum. We thusopted for active temperature control with limited proton beam currents. Theonly cooling provided in these experiments was via conduction to a smallaluminum support and convection to the surrounding air. In the beam transporttunnel, a jet of air was aimed at the target to enhance convective transfer.
Thermocouples were attached to each target, and we agreed on alimit of 300 C. In all cases, we were easily able to maintain thetemperatures below this limit by simply adjusting the beam current. At 50 MeVand a beam current of 1 pA (50 W), the maximum temperature in the aluminumtarget was 285 C. For the uranium target, the beam current had to be reducedto about 0.8 A (40 W) to keep the temperature below 310*C. At 113 MeV, abeam current of 4.8 A (542 W) produced a temperature of only 2800C inaluminum. At 256 MeV we had to periodically reduce the repetition rate to15 Hz to keep the temperature below 300 C. In all cases the initialtemperature rises were about 15 C/min for aluminum and 60C/min for uranium,in good agreement with the heat capacities. However, these steep rates onlylasted a few minutes, and the temperatures stabilized in about 10-15 min.When the beam was turned off or reduced, the temperatures began to fallimmediately, as expected, since the decay heat was trivial. Consequently,there was never any problem in controlling the temperatures, and there wasalways ample time to make changes in the beam current to avoid excessiveheating. Hence, no problem is expected in performing similar measurementswith uranium and lead in the future.
Further experiments are planned with lead and graphite targetsat 50 MeV and with lead, graphite, and uranium targets at 113 and 256 MeV. Wealso plan to irradiate foil'; at Michigan State University in February. Thesemeasurements will be conducted with a deuteron beam at about 120 MeV. On allcases, we will be able to compare our results with time-of-flight measurementsand High Energy Transport Code (HETC) calculations in order to demonstrate theutility of the technique.
A report has been comleted summarizing these experiments andhas been submitted to the Air Force Weapons Laboratory for final approval.
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55
II. SEPARATION SCIENCE AND TECHNOLOGY(G. F. Vandegrift)
The Division's work in separation science and technology consists ofthree projects. The first is concerned with developing a technology base forthe TRUEX (TRansUranic EXtraction) solvent extraction process. The TRUEXprocess extracts, separates, and recovers TRU elements from solutions con-taining a wide range of nitric acid and nitrate salt concentrations. Theextractant found most satisfactory for the TRUEX process is octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide, which is abbreviated CMPO. Thisextractant is combined with tributyl phosphate (TBP) and a diluent to formu-late the TRUEX process solvent. The diluent is typically a normal paraffinichydrocarbon (NPH) or a nonflammable chlorocarbon such as carbon tetrachloride(CC14) or tetrachloroethylene (TCE). The TRUEX flowsheet includes a multi-stage extraction/scrub section that recovers and purifies the TRU elementsfrom the waste stream and multistage strip sections that separate TRU elementsfrom each other and the solvent. Our current work is focused on facilitatingthe implementation of TRUEX processing of TRU-containing waste and high-leveldefense waste, where such processing can be of financial and operationaladvantage to the DOE community.
The major effort in TRUEX technology-base development involves developinga generic data base and modeling capability for the TRUEX process, referred toas the Generic TRUEX Model (GTM). The GTM will be directly useful for site-specific flowsheet development directed to (1) establishing a TRUEX processfor specific waste streams, (2) assessing the economic and facility require-ments for installing the process, and (3) improving, monitoring, andcontrolling on-line TRUEX processes. The GTM is composed of three sectionsthat are linked together and executed by HyperCard and Excel software. Theheart of the model is the SASSE (Spreadsheet Algorithm for Stagewise SolventExtraction) code, which calculates multistaged, countercurrent flowsheetsbased on distribution ratios calculated in the SASPE (Spreadsheet Algorithmsfor Speciation and Partitioning Equilibria) section. The third section of theGTM, the SPACE (Size of Plant and Cost Estimation) section, estimates thespace and cost requirements for installing a specific TRUEX process in a glovebox, shielded-cell, or canyon facility. The development of centrifugalcontactors for feed- and site-specific applications is also an important partof the effort. Sections II.A to II.I report on all aspects of the TRUEXTechnology-Base Development Project.
The objective of the second project is the development of a TBP-basedsolvent extraction step for recovering plutonium from pyrochemical saltresidues and a CMPO-based step to make the raffinate nonTRU. This effort isreported in Sec. II.J.
The objective of the third project is to determine the feasibility ofsubstituting low-enriched uranium for the high-enriched uranium currently usedin producing fission-product 9 9 Mo. Technetium-99m, the daughter of 9 9 Mo, iswidely used in medical diagnosis. This project is reported in Sec. II.K.
A. Development of Extraction-Behavior Models for the GTM
Mathematical models of equilibrium and kinetic extraction data for theGTM continue to be improved as more data are collected. The correlations are
56
based on chemical mass action principles in which the effects of organic phasespeciation, metal complexation, and aqueous phase activity coefficients areconsidered. In mixed electrolyte solutions, the stoichiometric activitycoefficients of H+ and NO3 are calculated by the method of Bromley.' Wateractivities of multicomponent electrolyte solutions are calculated by themethod of Kusik and Meissner.2 The details of these calculations have beenpresented elsewhere.3 '4 Organic phase species are treated in terms of theideal associated solution theory.*'8 The ideal approximation assumes that allorganic phase species behave ideally and that complex formation can be used toexplain any solution deviations from ideality. Physical interactions areassumed to be negligible in comparison with the chemical contribution tononideality.
1. Modeling of Nitric Acid Extraction(D. J. Chaiko)
a. Extraction by CMPO
Nitric acid extraction data for CMPO-TCE solutions have beenmodeled using a two-parameter model that has been described previously.3 Theextraction equilibria were defined as follows:
KC1
H+ + N03 + CMPOG CMPO'HNO3 (II-1)
KC2
2H+ + 2N0 + CMPOG_-CMPOG2HN03 (11-2)
In the above equations (and ones that follow), the overbar signifies anorganic phase species. Fitting this model to extraction data at 0.25, 0.50,and 1.OM CMPO, however, has shown that KC1 increases slightly with totalorganic phase CMPO concentration (see Table II-1 and Fig. II-1). Thisbehavior indicates that KC1 contains a term that includes CMPO concentration.This situation could result from the formation of a dimeric CMPO species asfollows:
KC3
H+ + NO + 2CMPO <_ (CMPO) 2aHN0 3 (11-3)
Table II-1. Equilibrium Constants for HNG3Extraction by the CMPG-TCE Solventat 250C
Equilibrium Constant[CMPO],
M KC1 KC2
0.25 1.9 0.1 0.009 0.004
0.50 2.21 0.07 0.014 0.004
1.0 2.5 0.2 0.012 0.006
57
3.0
2.5 F
2.0
U
1.01-
0.51-
0.0 L0.0 0.5 1.0 1.5 2.0
[CMPO], M
Fig. II-1. Plot of KC1 vs. Total CMPO Concentration at250C. The constant KC1 is obtained fromfitting an acid-extraction model based onEqs. II-1 and 11-2 with experimental data.
According to the equilibria in Eqs. II-1 and 11-3, at low acid concentrationwhere Eq. 11-2 is not significant, the total organic phase nitric acidconcentration would be given by
[HNO3]totai = KC1'{H+}{NO } [CMPO]free + KC3' {H+} [N03} [CMPO]free (II-4)
where the braces refer to activities and the brackets to molar concentrations.If the total organic phase nitric acid concentration is low, [CMPO]totai canbe substituted for [CMPO]free, giving:
[HN03]totaj = [CMPO]totai {H+}{NO3} (KC1' + KC3' [CMPO]totai) (II-5)
An extraction model based only on Eqs. II-1 and 11-2 gives an equilibriumconstant KC1, which is described by
KC1 = KC1' + KC3' [CMPO]t0tai (II-6)
This means that plotting KC1 vs.extraction data were obtained atFig. II-1 appears to confirm the[i.e., (CMPO)2 .HN0].-
[CMPO] should give a straight line. Althoughonly three CMPO concentrations, the graph inpresence of a dimeric CMPO species
- - - - , - - - - , . . . . , . . .
-
1 l 1
58
The extraction data for 0.25M and 1.0M CMPO were refitted usingan extraction model based on Eqs. II-1, II-2, and II-3. The data for 0.5MCMPO were not used in the curve fitting because they did not cover the regionof aqueous [HNO3] below 0.2M, where Eq. 11-3 has the most influence on organicHNO3 concentration. The resultant equilibrium constants are given inTable 11-2. Also listed in Table 11-2 are the equilibrium constants obtainedby fitting all the extraction data for 0.25M and 1.OM CMPO simultaneously.The variation in KCI is now less than 2%, whereas in the two-parameter model,this constant varied by approximately 32%. Thus, in going to a three-parameter model, a considerable improvement in the stability of KC1 and KC2with respect to [CMPO]t0 ta1 was achieved. The variation in KC3, however, was20% but that is not regarded as excessive when considering that the error inthe experimental data at extremely low nitric acid concentrations can be ashigh as 100%. The calculated and experimental extraction isotherms areplotted in Fig. 11-2. All three calculated isotherms were obtained usingKC1=1.60, KC2=0.010, and KC3=1.66.
Table 11-2. Equilibrium Constants for a Three-ParameterAcid-Extraction Model at 250C
Equilibrium Constants[CMPO],
M KC1 KC2 KC3
0.25 1.62 0.06 0.009 0.002 2.013 0.009
1.0 1.59 0.05 0.010 0.001 1.67 0.09
0.25, 1.0a 1.60 0.04 0.010 0.001 1.66 0.06
All data for 0.25 and 1.OM CMPO were fitted simultaneously.
101
C),
Cu
100
10-
10' 2
10 '
10-10- 3 1 0 2 -10.1 1 00 101 102
Fig. 11-2.
Aqueous [HNO3J, M
Experimental and Calculated Nitric AcidExtraction Isotherms for CMPO-TCE at 250C
1M CMPO
0.5M
0.25M
0
- - ----- - - -- - - --- - -------- - - -----
59
The fact that the extraction constants KC1, KC2, and KC3 arenot dependent on CMPO concentration suggests that the ideal associatedsolution approach is reasonable. In additional support for this approach,modeling of the HNO3-TBP-NPH system has shown that the acid extractionconstants in that system are independent of extractant concentrations.4
b. Extraction by TBP
A study of the influence of TBP concentration on the extractionof HNO3 by TBP-TCE solutions was undertaken. As in the case of CMPO in TCE,the formation constants for the TBP-HNO3 species [(TBP)2 *HNO3 and TBPOHNO3 ]were independent of TBP concentration for the two concentrations examined(0.75 and 1.OM TBP). It should be noted that, for the nitric acid concen-trations studied ( 7.OM), it was not possible to obtain a reliable estimatefor the formation constant of TBPe2HN03 . This species is only important at>8M HNO3 and, therefore, is not included in the present model. Curve fittingof extraction data produced the values of KT1 and KT2 listed in Table 11-3.Figure 11-3 illustrates the kind of fit that is obtained between the model andthe extraction data.
Table 11-3. Extraction Constants for the HNO3/TBP-TCESystem at 25 C
[TBP],M KT1a KT2b
0.75 0.167 0.004 0.203 0.0091.0 0.15 0.02 0.21 0.02
aFormation constant for TBPeHN03 .bFormation constant for (TBP)2 .HNO3 .
100
101 1MTBP
0.75 M TBP
0 10-
O DATAL.CDEL
10
10' 10'1 100 101
AQ [HNO3], M
Fig. 11-3. Extraction of HNO3 by TBP-TCE Solutions at25 C. Data at both TBP concentrations aremodeled with KT1=0.157 and KT2=0.203.
60
c. Extraction by TRUEX-TCE
To model the extraction of HNO3 by the TRUEX-TCE solvent, theseparate models for HNO 3 extraction by TBP-TCE and CMPO-TCE were combinedwithout any change in the equilibrium constants. To fit the data at low HNO3concentrations, it was necessary to include a mixed complex (CMPO.HNO3 *TBP) inthe model. The extraction data for the HN0 3/TRUEX-TCE system could then bemodeled quite successfully using only one adjustable parameter (KM). Thevalues of all the extraction constants are listed in Table 11-4. Figure 11-4shows the fit of the model to the data.
Table 11-4. Extraction Constants for the
HN0 3/TRUEX-TCE System at 250C
Constant Value Species
KC1 1.6 CMPOeHN03
KC2 0.01 CMPO*2HN0 3
KC3 1.66 (CMPO)2 eHN03
KMa 1.91 CMPO*HN0O3 TBP
KT1 0.157 TBPeHN03
KT2 0.203 (TBP) 2 HN03
aValue of KM determined from a least-square
analysis.
2. Modeling of Americium Extraction(D. J. Chaiko)
In modeling the extraction of americium nitrate from aqueous solu-tions of high ionic strength, it is necessary to account, explicitly, for theinfluence of water activity (a) on the extraction equilibria. This will alsobe necessary for any other metal salts which undergo partial or completedehydration during transfer into the organic phase.
The activity coefficients reported in the literature, such as thosefound in Robinson and Stokes,7 almost always refer to the unhydrated electro-lyte. This is due to the extreme difficulty in determining the hydrationnumber (h). Different types of experimental techniques give different valuesfor h, which arise primarily from the question of whether or not both theprimary and secondary hydration shells are influencing the experimentalmeasurement.
In the case of highly hydrated electrolytes, the hydration numberwill not be constant. It will decrease at high electrolyte concentrations,when the water to electrolyte concentration ratio becomes less than thehydration number. For the modeling work described here, it has been assumedthat the hydration number of Am3 + remains constant regardless of the ionicstrength.
61
.25M CMPO, 1.0M TBP-TCE
100
AO [HNO3I, M
0.25M CMPO, 0.75M TBP-TCE
100
AO [HNO3], M
0.25M CMPO, 0.25M TBP-TCE
1 00AD [HNO3J, M
Fig. II-4. Extraction of HNO3 by the CMPO-TCESolvent at Various TBP Concentrationsat 250C. Extraction constants arelisted in Table 11-4.
100
0z
0
10
1001 D
O DATA-MODE
101
100
Cob
0z
0
10' 1
- -. 0-0
0 DATA....-uon
10 41 0' ~
100
110
1)
0zs
10' 1
1 01
J i i A 1 1
i i 1 i si
0 DATA
1n"02IV''
L
62
Americium nitrate extraction at low acid concentrations can bedescribed by
K0Am(H20)9* + 3N03 + 3CMPO< Am(N03)3 (CMPO)3 + 9H20 (11-7)
In the organic phase complex, CMPO is assumed to be a mono-dentate ligand,while nitrate is considered bidendate.
During the aqueous phase complexation of Am3 + by nitrate, it hasbeen assumed that one water of hydration is lost for each bound nitrate, asfollows:
Am(H20) 9 + nN0& -< Am(H20)9-n(N03 )n-n + nH2O (11-8)
Based on Eqs. 11-7 and 11-8, the expression for DAm is
Ko [CMPO]free {N0}3 (DAm =InI9 I + {N05}" 9+ E Qn {H20}n{H20}
The constants KO and p6 can be obtained from a plot of the following
polynomial equation:
[CMP0]free {N0}3 _ 1 fi {N03} P2 {NO}2- -- + -- + . . . (II-10)
DAm {H20}9 Ko0 Ko {H20} Ko {H20}
The activity coefficients of the tracer-level americium species are taken as aconstant and are, therefore, included in the ,O's.
For a tracer-level component such as Am3 +, where concentrations are10-3 M, and with a mixed electrolyte solution at a total ionic strength 0.1,
the cross-differentiation equation8 gives:
8 In 7Am+ 8 In 7N3+ 0 (II-11)
a mNO 8 mAm3
where 7 is the activity coefficient, and m is the molal concentration for thespecies indicated in subscript. In a similar manner, the activity coefficientof Am3+ is also independent of hydrogen ion concentration.
The coefficients in Eq. II-10 were determined using Am3 + extractiondata for 0.25M CMPO-TCE from HNO3 and HNO3/NaNO3 solutions. Free CMPOconcentrations were calculated from the acid extraction model described above.The straight-line fit in Fig. 11-5 indicates that, at the nitrate concentra-tions studied ( 5M), the terms beyond Q1 are not significant. From the slope
63
14
In0
0
r0.O)
12
10
8
6
4
21r
0.0 1 2 3
{NO)I(OH2 0}
Fig. 11-5. Straight-Line Fit to AmericiumData from HNO3/NaNO3 Solutions
Extractionat 250C
and intercept, the values of KO and Pi were found to be 5.6 x 10 and 28.8,respectively.
The present analysis of nitrate complexation is somewhat differentfrom that reported earlier.3 At that time, values of Pi and Pa were found tobe 86.5 and 63.2, respectively. Since tracer-level Am concentrations wereused in the distribution measurements (organic phase [Am3 +] was estimated tobe 10-5 M from HN03-only solutions), small amounts of organic phase impuri-ties, especially acidic extractants, can have a significant influence on DAm.This is demonstrated in
[Am]cMpo + E [Am]EXDAm [Am] (II-12)
where the summation (E [Am] Ex) is taken for all the organic phase impuritiescapable of extracting americium. The high nitrate salt data in Fig. 11-5 areconsidered to be more reliable than data for Am extraction from HN03-onlysolutions because the presence of_inextractablenitrate salts enhances theextraction of americium, giving [Am]CMPO x1-[Am]E.
At high nitric acid concentrations (i.e., >1M), spectroscopicstudies have shown that nitric acid is associated with the organic phase metalcomplex through hydrogen bonding to the carbonyl group of CMPO.9 Inprinciple, each of the three CMPO molecules per americium is capable ofextracting nitric acid. This gives rise to the following set of organic phasemetal species:
[Am(N03)3 (CMPO)3 (HNO3 ).]3
EOn=0
0
8
00
w
[AD] ts..i = (II-13)
64
The expression for DAm then becomes
DAm =
[CMPO]Iree {N03}3 (Ko + K1{H}{NO3} + K2{H+}2 {N0}2 + K3 {H+}{N0}3]
[1 + pi fNOai } {H2 0}(II-14)
The constants K1 , K2 , and K3 were obtained irom curve fitting americiumextraction data from nitric-acid-only solutions by O.25M CMPO-TCE. The valuesof the extraction constants were 7.2 x 105, 2 x 103, and 9, respectively. Themodel, as depicted in Eq. 11-14, is capable of predicting Am extraction fromHNO3/NaNQ3 solutions, as well as from HNO3-only solution with a single set ofconstants. Figure 11-6 shows the fit of Eq. 11-14 to the extraction data forO.25M CMPO-TCE. The americium distribution ratio at 0.O1M HN03 was determinedby others10 after pretreatment of the organic phase with diazomethane. ThisprG.:edure results in the methylation of any acidic impurities that wouldotherwise be capable of extracting Am3 +.
4.4 M NaNO3
2M
1M
OM
1 0-3 1 io- 10' 100
Fig. 11-6.
Americium Extraction from HNO3 /NaNO3Solutions by 0.25M CMPO-TCE at 25*C.Lines calculated from Eq. 11-13.
101
Aqueous I[NO31,M
B. Estimating Densities of Complex Aqueous Solutions(D. B. Chamberlain, A. B. La'O, R. A. Leonard, and I. R. Tasker)
1. Density Correlation
In the GTM, molar concentrations (the scale commonly used todescribe feed solutions) are converted to molality (the scale upon which mostof the solution physical chemistry is based). For this conversion, thedensity of aqueous solutions is needed. This density correlation is derivedas follows.
I0
oC0
U,0
E4
102
101
100
10.2
10-__-- ... 1 . _ ..
t...
1
65
At constant temperature and pressure, by definition, the apparentmolar volume cf a solute is characterized by
V - VH2 OV, 2 =
where
V - n1V1
n2(II-15)
V = total volume of solution (mL)
VH2O = ni 1 = volume of solvent water (mL)
ni = moles of water, component 1
n2 = moles of solute, component 2
V1 = molar volume of water (mL/mol)
VO,a = apparent molar volume of solute (mL/mol)
On the molar concentration scale, by definition
V = 1000 mL (1I-16
Therefore, the moles of the solute in 1000 mL of solution equals the molarityof the solute in solution, or
n2 = c2 (II-17
)
)
where c2 is the molarity of the solute in solution.mass/volume, the mass of water in solution is equal[(1000 mL)d - mass of the solute]. Therefore,
o Q1000d - caMaVIR 2 p= ni~l = -
do
where M2 = molar mass of the solute (g/mol)
d = density of solution (g/mL)
do = density of pure water (g/mL)
Substituting Eqs. 11-16, 11-17, and 11-18rearranging gives
Since density =to VH2OdO and to
(II-18)
into Eq. 11-15 and
d = d0 +(M2 - [do x V ,2]) x ca
1000(II-19)
Hence, prediction of a solution density is a problem of obtaining theappropriate V#,2 . Our correlation involves two steps. First, for thepurposes of our work, we assume that the value of V6 ,2 under all solutionconditions may be adequately approximated by its infinite dilution value
66
V f,2 . (See below for further discussion.) Secondly, at infinite dilution,we adopt the additivity principle; for an electrolyte MX,
V MX v ++V,x (11-20)
Substituting Eq. 11-20 into Eq. 11-19 gives
(MM+ - [d x V +]) x c + (M - - [d x V j-]) x c -
d = d + 0+X(11-21)
1000 1000
Equation 11-21 may be readily extended to a multicomponent ionic solution togive the general equation
E (Mi - [do x VO,1]) x cid = do+1000 (11-22)
1000
This equation was included in the Generic TRUEX Model. A collection (database) of apparent molar volumes (AMV) for ions relevant to the chemistry ofthe process is given in Table 11-5. Most of these values were obtained fromthe publications of Millero.11,12 Apparent molar volumes for the metals Nit+,Cue+, Cd2 +, and Ca2+ were obtained from LoSurdo and Millero,1 3 while those forNa+ and NO3 were obtained from Roux,1 4 who presented data based on theoriginal data of Millero.
As noted by the table footnotes, not all of the apparent molarvolumes given in Table 11-5 are values obtained from the literature. Forthose ions, the partial molar volume values have been estimated in thefollowing manner.
(1) The apparent molar volume of Ce3 + is the numeric averageof the values for La3 + and Pr3 +, its two neighboring elements. This procedurewas also used for apparent molar volumes of Pm3 + (average of Nd3 + and Sm3 +)and Eu2 +(average of Sm3 + and Gd3 +).
(2) The apparent molar volume of Y3 + was determined byassuming that the ratio of V#(Sr2+)/V#(Ba2 +) is the same as the ratio of
VO(Y3+)/VO(La3+), where Sr2 +, Bat+, and La3+ have published apparent molar
volume values.
(3) The apparent molar volumes of C104-, Zr4+, and HC 204- wereall estimated using experimental data. For example, densities of ZrF4solutions with various concentrations of Zr4+ and F- were measured with adensity meter. Since the apparent molar volume of F- is known, the apparentmolar volume of Zr4+ could then be calculated from Eq. 11-22. This procedurewas also used for determining the apparent molar volumes of C104 ~ (using HC104solutions) and HC204 (using H2C204 , HN03 solutions). The C104- ion does notappear in the GTM, but its apparent molar volume was needed to calculate somedata used in modeling. An experimental estimate of the apparent molar volumewas used as an expedient since the use of the literature value at infinitedilution led to large errors in density predictions for solutions of finiteconcentrations.
67
Table 11-5. List of Apparent Molar Volumes Used in the GTM
Cation V#,2 Comment Cation VO,2 Comment Anion V#,2 Comment
Ag+ -0.7 Nd3 + -43.31 C104- 22.83 g
Al3+ -45.3 Ni2+ -29.5 F- -1.16
Am3+ -42.33 a Np4+ 0 j HC204- 49.15 g
Ba2 + -12.47 NpO2 + 0 j H2 PO4- 29.1
Ca2+ -17.8 Pd2+ 0 j HS04~ 35.67
Cd2 + -14.2 Pm3+ -42.82 d N03 - 29.5
Ce3 + -40.82 b Pr3+ -42.53 Tc04- 22.83 h
Cm3 + 0 j Pu3 + 0 j C1 18.1
Cr3 ; -39.5 Pu4 + 0 j
Cs* 21.34 Rb+ 14.07
Cu2+ 25.5 Rh2 + 0 j
Eu3 + -41.37 c RuNO3 + 0 j
Fe3 + -43.7 Sm3 + -42.33
Gd3 + -40.41 Sr2 + -18.16
H+ 0 i Tc4+ 0 j
La3+ -39.1 U0 22+ 15.91 e
Mg2+ -21.17 Y3+ -56.94 f
Na+ -1.2 Zr"* 115.94 g
aEstimated value (assumed same AMV as Sm3+).
bEstimated value (average of La3+ and Pr3+ AMY's).
cEstimated value (average of Sm3 + and Gd3+ AMV's).
dEstimated value (average of Nd3 + and Sm3 + AMV's).
'Estimated value using published density data.
IEstimated value (assumed ratio Sr2 +/Ba2+ AMY's same as Y3 +/La3 + ratio).
gEstimated value based upon experimental density data.
hEstimated value (assumed same AMY as C104-).'H+ ion set as reference AMV.
iAMV's not available for this component. Because the concentration of thiscomponent is very low, it will not affect solution densities. Therefore,AMY set equal to 0.
68
(4) The apparent molar volume of UOa+ was estimated frompublished density data for various solutions of U02(N0 3)2 .15 Theconcentrations and solution densities, along with the apparent molar volume ofNOB, were used to determine the apparent molar volume of U02+ (fromEq. 11-22).
(5) The apparent molar volume of Tc04- was assumed to be equalto that of 0104-.
(6) The apparent molar volume of Am3 + was assumed to be equalto that of Sm3 +.
For cations whose apparent molar volume could not be determinedeither experimentally or through the literature, a value of 0 was specified.These ions include Cm3 +, Np4 +, NpO2 +, Pd
2+, Pu3 +, Pu4 +, Rho+, and RuNO3 +.
Regarding the derivation of the density correlation, acriticism that might be leveled is the identification of V 0 ,2 with V, 2 . Itis true that, in the most rigorous work, recognition must be taken of theconcentrative behavior of VO; much work has been directed toward this goal,and relations such as the Masson equation'8 are available:
V0 = V + S*C1/2 (11-23)
When incorporating this equation into Eq. 11-19, it yields the Root Equation' 7
(M2 - [do x Vt]) x c SZ x do x c3 "2
d = do +- (11-24)1000 1000
where the S* term has indications of additivity.'2 However, S* data are notavailable for most ions of interest to this work, and although one may makeestimates of them using the Debye-Huckel approach, the contribution of thefinal term in Eq. 11-24 is usually of the order of only a few tenths of onepercent in density. An error of this magnitude is significant in workdirected toward gaining a theoretical understanding of volumetric behavior(e.g., 0.1% error in d can cause several thousand percent error in Vi), but interms of the conversion between molarity and molality, the error will usuallybe far less than 1%, a level insignificant for the present purposes.
2. Density Correlation vs. Measured Density Data
A comparison of the measured and calculated densities for a numberof different solutions is reported in Tables 11-6 through II-10. Data fromthese tables plus additional density measurements are summarized inTable II-11. Based upon these data, the density correlation typicallypredicts the measured density to 3%.
69
Table 11-6. Measured and Calculated Densities for HNO3-NaNO3Solutions
Conc., mol/L Density, g/mL
[NO3] [Na+] Measured Calculated % Error
5.5
3.256
4.209
3
2
6.98
6.48
5.98
4
5
1.365
3.765
4.635
5.488
7.225
1.5
2.5
4.9
2
3
5.4
3
4
6.4
5
6
67
7
8
9
2.02
2.05
2.1
2.5
3
4
8
4.5
3
4
2
1
5
4.5
4
3
3
1
23
4
61
2
4.4
1
2
4.4
1
2
4.4
1
21
21
21
2
22
2
2
2
2
1.259
1.162
1.211
1.134
1.085
1.309
1.286
1.262
1.184
1.213
1.063
1.161
1.208
1.252
1.341
1.073
1.121
1.242
1.085
1.137
1.260
1.119
1.1751.301
1.183
1.229
1.210
1.263
1.241
1.287
1.292
1.106
1.1073
1.1094
1.1208
1.1362
1.1675
1.2864
1.285
1.176
1.231
1.143
1.086
1.345
1.317
1.289
1.200
1.233
1.066
1.168
1.221
1.273
1.378
1.070
1.127
1.263
1.086
1.143
1.279
1.119
1.176
1.3121.1841.2411.217
1.274
1.249
1.306
1.315
1.111
1.112
1.114
1.127
1.143
1.176
1.306
2.03
1.18
1.65
0.79
0 14
2.74
2.37
2.07
1.31
1.57
0.300.63
1.06
1.60
2.69
-0.25
0.50
1.67
0.09
0.57
1.52
-0.03
0.05
0.83
0.13
0.94
0.52
0.830.65
1.49
1.740.47
0.44
0.40
0.54
0.61
0.70
1.51
70
Table 11-7. Measured andSolutions
Calculated Densities for HNO3-HC104
Conc., mol/L Density, g/mL
[N03] [C104] Measured Calculated % Error
0.0 1.047 1.0827 1.078 -0.50
0.2 0.838 1.0725 1.068 -0.44
0.5 0.523 1.0666 1.054 -1.25
0.7 0.314 1.0570 1.044 -1.25
Table 11-8. Measured and Calculated Densities for 101-10104
Solutions
Conc., M Density, g/mL
[Cl-] [C104] Measured Calculated % Error
0.491 3.664 1.2789 1.288 0.61
0.887 3.141 1.2505 1.255 0.25
1.965 2.094 1.1882 1.193 0.35
2.947 1.046 1.1283 1.130 0.09
3.439 0.523 1.0996 1.098 -0.15
4 0 1.066 1.068 0.16
Table 11-9. Measured and Calculated Densities for HN03 -Al(N03 )3Solutions
Conc., mol/L Density, g/mL
[N03] [A13 +] Measured Calculated % Error
2.03 0.67 1.1068 1.112 0.24
2.06 0.67 1.1073 1.113 0.28
2.11 0.67 1.1083 1.114 0.34
2.51 0.67 1.1194 1.127 0.51
3.01 0.67 1.1334 1.143 0.70
4.01 0.67 1.1616 1.176 1.05
8.01 0.67 1.2703 1.306 2.62
71
Table II-10. Measured and Calculated Densities for HNO3-NaNO3 -Al(NO3)3 Solutions
Conc., mol/L Density, g/mL
[N03] [Nat] [Al 3+] Measured Calculated % Error
5 1.6 0.8 1.2363 1.256 1.615 1.2 0.6 1.2229 1.232 0.765 0.8 0.4 1.1972 1.208 0.915 0.4 0.2 1.1294 1.184 4.62
Table II-11. Measured and Calculated Densities for Various Solutions
Density, g/mLNo. of Range of Molar Range of
Solution Ions Concentration Meas. Calc. Error, %
PFPa 17 3.07M NO 1.1341 1.142 0.72
0.44M Al3 +
PFPb 8 3.06M NOS 1.1350 1.147 1.1
0.43M Al3
CAWC 40 3.96M NO 1.2663 1.286 1.5
CAWd 40 3.55M NO 1.2046 1.218 1.1
CAW* 11 1.19M NO 3 1.1853 1.208 1.9
HNO3-NaNO3 3 1.365-9M NOS see Table 11-6 -0.25 to 2.69
1.0-6M Na+
HNO3 -HC104 3 0-0.7M NO see Table 11-7 -2.9 to -2.010.3-1.OM C10~
HCl-HC104 3 0.5-4M Cl see Table 11-8 -5.8 to 0.16E0-3.7M C104
HNO3-Al(N03)3 3 2.03-8.01M NO see Table 11-9 0.33 to 2 .9g0.67M Al3
HNO3-NaNO3-Al(N03 )3 4 5.OM NO see Table 11-10 1.6 to 4 .8h
0.4-1.6M Na*
0.2-0.8M Al3 +
Simulated Plutonium Finishing Plant Waste Stream, G. F. Vandegrift.bSimulated Plutonium Finishing Plant Waste Stream, D. J. Chaiko.cSimulated Current Acid Waste Stream, twice typical concentration,
N. Simonzadeh.dSimulated Current Acid Waste Stream, typical concentration, N. Simonzadeh.'Simulated Simplified Current Acid Waste Stream, A. B. La'O.TError increases with increasing [C104].
gError increases with increasing [NO].hError of 4.8% corresponds to 5M NO3, 0.4M Na+, 0.2M Al3 +.
72
C. SASSE Development(R. A. Leonard, D. B. Chamberlain, J. M. Leddin, and W. B. Seefeldt)
An electronic worksheet called SASSE (Spreadsheet Algorithm for StagewiseSolvent Extraction) has been developed to allow detailed evaluation ofproposed flowsheets in conjunction with information from the GTM data base(see Sec. II.D). In addition to establishing that each effluent will be ableto reach its design composition, the SASSE worksheet (spreadsheet) can be usedto identify key points for process monitoring and control. An early versionof SASSE was used to assist the Westinghouse Hanford Co. in countercurrenttesting of the TRUEX flowsheet using actual Plutonium Finishing Plant (PFP)wastes. An Excel macro has now been written that will generate a new SASSEspreadsheet for a given number of (1) sections, (2) stages in each section,and (3) components. A second Excel macro has been written that makes it easyto modify both the number of stages in a section and the number of sections inthis SASSE spreadsheet. Both macros need to be tested in detail and thenproperly documented so that others can easily generate, use, and modify SASSEworksheets. Finally, the output from the SASSE worksheet will be combinedwith plant-specific information in a separate worksheet to calculate plantsize and capital costs.
The macro to generate a SASSE worksheet was used for verification test 1(Sec. II.G). The macro worked quite well and, with only a few changes, wasready for calculating the stage-to-stage compositions. Since the GTM database is not ready, simple empirical correlations were used for calculating thedistribution coefficients of nitric acid and neodymium. Based on thesecalculations, the aqueous flow rates to the scrub sections were increased, andthe nitric acid concentration in these feeds was reduced, so that the nitricacid concentrations in the organic phase going to the first strip would be lowenough to allow most (95%) of the neodymium to be recovered in the aqueouseffluent from the first strip.
Work was started on a HyperCard stack that will act as a front end to theGTM, which includes SASSE. The use of an Excel macro to generate the SASSEworksheet will be one of the keys to making the GTM work. The dialog boxesnow used when SASSE is generated independently will be replaced by informationobtained from the HyperCard stack.
A general equation was developed for the stage-to-stage calculations. Itmight be used in further revisions of SASSE as it allows for either theaqueous or organic phase or both to have an external feed stream at eachstage. In addition, it allows the effluent for either the aqueous or organicphase or both to be partially or totally returned to the next stage as aninternal feed stream. This general equation follows from the schematic forstage i shown in Fig. 11-7.
In addition, the organic-to-aqueous flow ratio (Ri) for the stage isdefined as
q0 , iRi = (11-25)
qa,i
73
f0,1 q0 -
q a, I
XI
yoI, fY11f
yI
Stage i
fal q aI+1Xi+1
ge,i,f
Xi, f
Fig. 11-7. Schematic for Stage iwith External and InternalFeeds for Both Phases
f~ i = fraction of phase # (either a, aqueous, or o, organic) from theadjacent stage (either i + 1 or i - 1) that flows to stage i. Ifthis fraction is greater than zero, then there is an internalfeed of phase 0 from the adjacent stage.
qoi = flow rate of phase 0 (either a or o) from stage i, L/min.
q , ,f = flow rate of external feed of phase # (either a or o) to stage i,L/min.
xi = concentration of component j in aqueous effluent from stage i, M.
xi,f = concentration of component j in the external aqueous feed tostage i, M.
yi = concentration of component j in organic effluent from stage i, M.
yi f = concentration of component j in the external organic feed tostage i, M.
74
And the stage is assumed to be an equilibrium stage so that the distributioncoefficient (Di) is given by
yiDi = - (11-26)
The flow rates for the two effluents from stage i are given by
qo,i = qo,i,f + foi qo,i_1 (11-27)
qai = qa,i,f + fa,i qa,i+l (11-28)
When these equations are combined with a material balance around the stage forcomponent j and solved for the concentration of component j in the aqueouseffluent from stage i, the result is
go,i,f qo,i-1 Ra,i,f qa,i+i
qo iyif + foi Riyi_1q+ij , f+ qfa,1 . xi+ 1
xi = 1 + RD+ (11-29)
This is the desired general equation for stage i. Note that it does not takeany account for other-phase carryover in either effluent from the stage.
Work was done on the the input that will be needed for the worksheet usedto calculate plant size and capital costs. Once the required input isidentified, work can proceed on this worksheet. In this worksheet, it will beassumed that appropriate facilities already exist with the necessaryutilities, analytical services, plant protection, health physics services,etc. If such facilities do not exist or are inadequate, the worksheet usermust do additional calculations to determine the cost of building a newfacility or upgrading an existing facility. These calculations will not bedone by the worksheet for plant size and capital costs, but they will useinformation calculated by this worksheet.
Input from the SASSE worksheet to the worksheet for plant size andcapital costs includes (1) the number of sections, (2) the number of stages ineach section, (3) the maximum flow rate of the waste stream that is the feedto the TRUEX process, (4) the TRUEX solvent used, and (5) the relative flowrates of the other inlet and outlet streams. Site-specific information to theworksheet for plant size and capital costs includes (1) the maximum storage-tank volume required for each input and output stream that is fed to, orcollected from, the TRUEX process, (2) the storage tanks that must have acriticality-safe geometry, (3) the feeds and effluents that must be in a glovebox, and (4) the feeds and effluents that must be shielded and the type ofshielding that will be required.
75
Future work will involve (1) further testing of the two SASSE macros, (2)finding the best way to bring the SASPE correlations into SASSE as they arecollected, (3) developing an algorithm for calculating plant size and costsbased on the results of the SASSE worksheet, and (4) finding the best way tolink the SASSE worksheet with the size/cost algorithm.
D. Development of the Generic TRUEX Data Base(W. B. Seefeldt)
The success of the TRUEX process and the development of models useful indesigning flowsheets are strongly dependent on the quality and theretrievability of underlying measurements of distribution coefficients of themany chemical species likely to be present.
A program was initiated to collect all such information into a computerdata base using the software package 4th Dimension on a Macintosh computer.The design of the data base will place strong emphasis on various modes ofretrievability of the data and on anticipated report formats useful to theproject.
The data base is intended to be useful to at least four customer types:program managers, individual experimenters who measure distributioncoefficients, modelers who develop the algorithms needed for the design ofTRUEX flowsheets, and quality assurance auditors. The ability to rapidlyidentify categories of information in need of development is especiallyimportant to program managers and modelers.
The design of the data base has been initiated.
E. Measurements of the Extraction Behavior of Americium and Nitric Acid
(D. R. Fredrickson)
1. Effects of Nitrate Salts on Am Extraction
Three series of experiments have been run to study the effect ofvarious nitric salt concentrations on the extraction of 241 Am by 0.25M CMPOdiluted by TCE. In addition, three series of experiments have been run tostudy the effect of various nitric-acid/nitric-salt concentrations on theextraction of 241Am by TRUEX-NPH solvent (0.2M CMPO-1.4M TBP in ConocoC1 2-01 4). In this study, the CMPO was purified in our laboratory andconsidered to be 98% pure. Each of the series covered a nitric acid rangethat included 0.01, 0.1, 0.5, 1, 2, and 3 or 4M acid. The three NaNO3 saltconcentrations were 1, 2, and 4.4M. The experimental results are given inTables 11-12 and 11-13.
2. Nitric Acid Extraction
Two series of experiments have been run to study the extraction ofnitric acid by (a) 0.75M TBP in TCE and (b) 1.OM TBP in TCE. Each of the twoseries covered a nitric acid range from 0.05 to 7M (eight acids). Theequilibrium organic phase concentration of nitric acid was determined bytitration for all these measurements. The data are presented in Tables 11-14and 11-15.
76
Table II-12. 2 4 1Am Extraction by O.25M CMPO-TCE in Contactwith Various HNO3/NaNO3 Concentrations at 25 C
DAm
1M NaNO3 2M NaNO3 4.4M NaNO3[HNO 3],
M Forward Reverse Forward Reverse Forward Reverse
0.01 125 114 516 - 3803 -
0.1 114 102 332 335 1133 1207
0.5 63.3 61.9 90.7 94.5 95.3 97.2
1.0 30.1 28.9 30.5 32.9 22.6 22.6
2.0 8.64 8.04 7.55 7.80 4.75 4.61
3.0
4.0
2.47 2.35
2.23 2.01 1.90 1.95
Table 11-13. 241Am Extractions by TRUEX-NPH in Contact with VariousHNO 3/NaNO3 Concentrations at 250C
1M NaNO3
Forward Reverse
68.1 68.0
64.9 67.1
57.3 55.3
48.4 51.9
42.2 42.6
DAm
2M NaNO3
Forward Reverse
340 345
233 237
115 119
79.1 79.5
49.5 47.2
4.4M NaNO3
Forward Reverse
2927 3028
931 1142
231 238
122 122
54.4 54.4
33.5 33.5
27.5 27.7 26.2
[HN03 ],M
0.01
0.1
0.5
1.0
2.0
3.0
4.0 26.6
77
Table 11-14. Distribution of Nitric Acid betweenO.75M TBP/TCE and Aqueous NitricAcid Solutions at 25C
[HNO 3], M
Aqueous Phase
0.0506
0.1008
0.3073
0.5052
1.017
3.035
5.066
6.923
Organic Phase
0.0000460.000066
0.0011890.001175
0.011600.01130
0.028670.02884
0.1007
0.4360
0.6455
0.7273
Table 11-15. Distribution of Nitric Acid between1.OM TBP/TCE and Aqueous Nitric AcidSolutions at 25*C
[HN03], M
Aqueous Phase
0.0506
0.1008
0.3073
0.5052
1.017
3.035
5.066
6.923
Organic Phase
0.0021930.002152
0.018220.01845
0.045890.04584
0.1438
0.5698
0.8243
0.9209
DHNO 3
0.00090.0013
0.00120.0012
0.0380.037
0.0570.057
0.099
0.144
0.127
0.105
DHN0 3
0.0210.021
o .059
0.060
0.0910.091
0.141
0.188
0.163
0.133
78
F. Effects of CMPO and TBP Concentration on Extraction Behavior(P.-K. Tse)
1. Americium Extraction
In previous studies, we showed that CMPO is a more powerfulextractant than TBP for americium.1 8 The influence of TBP on americiumextraction by CMPO has 'een studied by Chiarizia and Horwitz.19 The TBP inthe TRUEX-process solvents can be described as: (1) improving phasecompatibility (reduces the likelihood of third-phase formation), (2) enhancingDA above 1M HNO3 , and (3) decreasing DA. below 1M HN03 . The effects of thevariation of both CMPO and TBP concentrations have not been studied and arethe focus of this work.
The distribution ratios of Am in 2.0 and 6.OM HNO3 as a function of[CMPO] and [TBP] are displayed in Figs. 11-8 and 11-9. Figure 11-8 shows thatthe slope of the DA vs. [CMPO] line decreases from 3.04 to 2.74 for 2M HN03and from 2.99 to 2.53 for 6.OM HN03 as the concentration of TBP increases from0.3 to 1.2M.
100
10
10
10~10 i
[CMPO], M
Fig. 11-8.
Americium Distribution Ratios as aFunction of [CMPO] for Various TBP in TCECompositions at 250C. (Open symbolsrepresent 2M HNO3 and closed symbolsrepresent 6M.)
TBP
o .3M
A~ .6M
o 1.2M
" .3M
a .6M
1.2W
0
00
C0
.0
U
4)
E
79
102
0
0~
0
10.
v
V
.1M CMPO.
A .3M CMPO
10 o0.5M CMPO
" .M CMPO
" . 3M CUPO
.5M CUPO
10 '
[TBP], M
Fig. 11-9. Americium Distribution Ratios as a Function of[TBP] for Various CMPO in TCE Compositions at250C. (Open symbols represent 2M HN03 and closedsymbols represent 6M.)
Similarly, Fig. 11-9 shows that the linear slope of D vs. [TBP] decreasesfrom 0.78 to 0.40 for 2M HNO3 and from 1.13 to 0.57 for 6.OM HN03 as theconcentration of CMPO increases from 0.1 to 0.5M.
2. Technetium Extraction
The effects of varying the [CMPO] and [TBP] in TCE were alsomeasured in the extraction of HTcO4 . The distribution ratics of technetium at250C as a function of [HNO3] with the standard TRUEX-TCE solvent and with0.25M CMPO and 0.75M TBP alone are given in Fig. II-10. The results indicatethat the combination of CMPO and TBP increases the technetium extraction overthat of the two separately at any nitric acid concentration. In the absenceof TBP, the DTc increases as [CMPO] 2 O (Fig. II-11). However, in the presenceof TBP, the stoichiometry of DTc vs. [CMPO] decreases from 2 to 1.15 as [TBP]increases from 0 to 1.2M (Fig. II-11). Similarly, the dependency of DTe on[TBP] decreases from 1.3 to 0.4 as the concentration of CMPO increases from0.1M to 0.5M (Fig. 11-12).
A CMPO-dependence experiment was performed to study the effectof CMPO on the extraction of technetium in the presence of 1.4M TBP in an NPHsolution. The results are shown in Fig. 11-13. As the concentration of CMPOis increased from 10-4 M to 10-2M, DTc is unaffected by its presence. However,between 0.1M and 0.4M CMPO, DTI was proportional to [CMPO].
10 -
0 id
10 -
0
E
c
.)
10-31C-2
Technetium Distribution Ratio for TRUEX-TCE, 0.25M CMPO,and 0.75M TBP in TCE vs. Nitric Acid Concentration at 25*C
0
C0
do
E
C4u
10
10
1 -j 0
0
x
without TBP
.3M TBP
.75M TBP
1.2M BP
10-2110 10
[CMPO]. M
Fig. II-11. Technetium Distribution Ratio vs. CMPOand TBP Concentration in TCE at 250C
80
C00, O 0.25M CMPO 0.75M TBP
t 0.25M CMPO
o 0.75M TBP _
0
J
10I
Fig. II-10.
10[HNo3], M
/
'slope = 2.0
--
. . A A I 1 9 1 a TRAIN rTl
81
- -
slope = 1.310
5 I I 61 11
100[TB], M
0
0
x
V
10
Fig. 11-12.
10 0t
10
Technetium Distribution Ratiovs. TBP and CMPO Concentra-tion in TCE at 25*C
-310
[CMPO], M
---
10
Fig. 11-13. Technetium Distribution Ratio between 1.0MNitric Acid and Various CMPO Concentrationsin 1.4M TBP Dissolved in NPH at 250C
1n
0
0~
C
0
.0
N_ 10-
E
C-tV
.1M CMPO
.2M CMPO
3M CMPO
.4M CMPO
.5M CMPO
10 410~
1o2
10i
0
C
.0
.')0
E
C
UGH
slope = 1.06
0O
100
I
82
G. Verification Studies(D. B. Chamberlain, K. A. Barnthouse, R. A. Benson, A. B. La'O,R. A. Leonard, and J. E. Stangel)
1. Introduction
Laboratory verification tests of the TRUEX process will be completedto (1) develop a better understanding of the chemistry of the TRUEX process,(2) test and verify process modifications and enhancements, and (3) verify theresults of the GTM being developed for predicting species' extraction behaviorand calculating flowsheets for the TRUEX process. These tests will be com-pleted in 4-cm centrifugal contactors during FY 1988, FY 1989, and FY 1990.
Three 4-cm centrifugal contactor units are available for completingthese tests. A sixteen-stage 4-cm centrifugal contactor located in a VacFrame Hood will be used for nonradioactive tests. A second sixteen-stage unitlocated in a once-through-air glove box will be used when radioactive isotopesare required. A new eight-stage 4-cm unit, designed for remote operation, canbe used for either radioactive or nonradioactive experiments.
Although specific waste solution compositions will be used in thesestudies, the purpose of these tests is not to demonstrate flowsheets forspecific waste streams, but to collect data to verify that the GTM predictsactual extraction behavior. Because these tests are not demonstration tests,the flowsheets that will be used have not been optimized.
A large number of samples will be collected in each of theseexperiments. Typical analyses of the nonradioactive samples will includemetals analysis (by inductively coupled plasma atomic emission spectrometry,ICP-AES), anion analysis (by ion chromatography, IC), acid analysis (by pHtitration), solution density measurements, and ion-specific electrode analysisfor NO and F- concentrations. For tests with radioactive components, theradioactive elements will be measured with appropriate analytical techniques.With these samples, other analyses will be limited because much of theequipment needed for detailed analysis is available only for nonradioactivesamples. Therefore, data collected in the nonradioactive tests will be usedto aid in the analysis of the data from the radioactive experiments. Toaccomplish this, conditions set for the nonradioactive tests will beduplicated in the radioactive experiments.
2. Future Plans
An outline of the verification tests planned for FY 1988 and FY 1989is listed below:
" Simulated Hanford Current Acid Waste (CAW) Solution--TRUEX-NPH" Simulated Hanford CAW Solution--TRUEX-TCEe Iron Extraction by both TRUEX-TCE and TRUEX-NPHe Solvent Carry-Over from Contactor Operatione Solvent Cleanup Tests" Tc04~ Extractions" Americium Recycle Test" Idaho High-Level Waste--TRUEX-NPH" Modified-PUREX/TRUEX Process
83
Other tests may be performed, since results obtained from one experiment maysuggest the need for new experiments or the elimination of others. Inaddition, several tests will probably be conducted using the same feedsolutions and the addition of different radiotracers.
3. Verification Run 1
a. Flowsheet
In verification Run 1, the extraction of neodymium from asimplified CAW solution using the TRUEX-NPH solvent (0.2M CMPO, 1.4M TBP inNPH) was investigated. Neodymium was used to simulate the extraction ofitself and other rare earths, as well as to model the extraction behavior ofamericium. In addition to the extraction of neodymium, the extractionbehavior of nitric acid, iron, oxalic acid, and fluoride were studied.
The flowsheet developed for this test is shown in Fig. 11-14.Two scrub sections were included; the first scrub was added to scrub nitric
TRUEX Sc(DX)
CMPOTBP 1
NPH dilue(200 m/mi
Feed (DF)
HNO3 1.19MNd 3+ 0.008MA13+ 0.72MFe 3+ 0.13MNa + 0.18MC204 2-0.18MF- 0.15MSO4 2- 0.27MNO3 - 3.26M(300 mb/min)
Strip #1 (FF)
HI 03 0.04MScrub #1 (Ds) (200 mb/mm)
HNO3 0.04M(5- nLmin) Scrub #2 (EF) Strip #2 (GF)
HN03 0.04M HF 0.1MFe3+ 0.005M (150 mUmin)
(100 mb/min)
k9 4'l 121314 116Ii
1
2 3 4 5 r6 7 1I1S1 1 3 4 5 1
Scrub #1
alhinate (DW) (DWI) Am Product (FW) Pu Product (GW)
HNO3 2.34M03 0.74 M (10 mdmin) HNO3 0.14M HN03 0.004M140 mUmin) (200 mb/mi) (150M jmin)
Scrub #2 Product(EW)
solvent Spent Sol1I1tHNO3 0.67M (GP)
(100 mmMin)2CMPO
0.2M TBP I
.4Mntn)
NPH dilu(200 mUm
Vent
0.2M1.4Mentin)
Fig. 11-14. Flowsheet for Actinide Removal from Synthetic CAW Waste
using the TRUEX-NPH Solvent, Verification Run 1
so-1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
l'u, -ow
R
HNC
(34
84
and oxalic acid from the organic phase. A small stream was collected fromstage number 6 (DW1) to analyze the composition of the aqueous phase in thescrub section. The second scrub (EF) contains Fe3+, which is used to stripoxalic acid from the organic phase. All of this stream was removed at stage 9so that its composition could be monitored throughout the run.
A simulated CAW solution was prepared for the feed to theextraction section. This CAW solution was prepared in a 20 L carboy andcontained 10 ionic species (excluding CO2-). The composition of thissolution is given in Table 11-16. Preparation of this solution consisted ofdissolving each water-soluble salt, including nitrates and sulfates, in waterbefore combining them in the carboy. Because neodymium carbonate wasdifficult to dissolve in water, nitric acid was added. Additional nitric acidwas added to make up for the acid used up from the reaction of Nd2 (C03)3 andHNO3 . Nitric, sulfuric, and hydrofluoric acids were added last. The solutionwas then brought up to 20 L with distilled water.
Table 11-16. Preparation of Simplified CAW Solution (ABL 492-104)
MolarSpecies Amount Added Concentration,Added Form of Additions to 20 L mol/L
H+ HNO3 , 12 0204 -2H20, HF, H2SO4 1.19b
N03 Fe(N03)3, Ca(N03)2, Al(N03 )3 , 3.26cHNO3 840.19 mL of 15.9M
C2042- H2C204-2H20 438.32 g 0.17
F- HF 6.92 mL of 28.9M 0.01NaF 117.57 g 0.14
total = 0.15
HS04- H2SO4 255.56 g of 18M 0.23NaHSO4 0.04
total = 0.27
Nd3+ Nd2 (C0 3)3 25.94 g 0.008
Fe3+ Fe(N03)3 -9H20 1050.45 g 0.13
Al3 + Al(N03 )3 -9H20 5401.87 g 0.72
Na+ NaF 117.57 g 0.14NaHSO4 110.48 g 0.04
total = 0.18
Ca2+ Ca(N03)2-4H20 94.46 g 0.02
the 20 L carboy.Nda(CO3 )3 in
GAll salts were dissolved in distilled water before adding tobThis value takes into account the acid used up by dissolvingHNO3 .CThis value is the total of all nitrate additions.
85
Eight liters of the TRUEX-NPH solvent was prepared for thistest. The CMPO used in this test was purchased from M&T Chemical. To preparethe solvent, the CMPO was purified, then mixed with the TBP and the NPH. Inadding the CMPO, it was assumed that the CMPO purity was 98%.
b. Test Results
Verification Run 1 was completed on March 4, 1988. Thechecklist used in this run is shown in Table 11-17 and includes both theplanned schedule and the actual times of the various events. As the tableshows, this run proceeded fairly close to the planned schedule until 22minutes and 15 seconds into the run. At this point, approximately two minuteshad been set aside to verify that the organic phase was not exiting with anyof the aqueous raffinates and to measure the various process flow rates. Atthe first flow-rate check (at 22:15), the second scrub raffinate (EW)contained a small fraction of organic phase in 100 mL of aqueous phase.Shortly thereafter, however, the three-way sample valve on the organic productstream (GP) was misaligned after a sample had been collected, which restrictedthe organic discharge from stage 16 and backed the organic phase up into thisstage. This organic material was then discharged in the GW raffinate(stage 15). During the next 10 minutes of the run, organic material backed upinto the last contactor stage on two additional occasions, both times due tothe slight misadjustment in the three-way sample valve. (The sample valve onthe GP stream had to be aligned perfectly in order to prevent the organic flowfrom being restricted and backing up into stage 16. This valve was eliminatedin later runs.)
Once problems with the GP sample valve were corrected, thesampling period was completed without additional incidents. At the end of therun, however, organic material was observed in the second scrub raffinate (EW)tank and both strip raffinates (FW and GW). Aqueous material was not observedin the organic product (GP), nor was organic material observed in the extrac-tion section raffinate (DW). As in the solvent cleanup test (completed beforethis run to clean up the organic solvent and train operators), organicmaterial was observed in one of the raffinate streams (GW). It appears thatthe contactors were being operated at flow rates beyond their capability. Theproblems with organic phase back-flowing into stage 16 also contributed to theorganic material in the second strip raffinate.
Solutions contained in the annular region of the contactors andin the rotors were collected separately for analysis. When stages 1 and 2were drained, more solution than expected drained out, overflowing the beakerplaced beneath these stages. After overflowing the beakers, liquid wasobserved in both high-level probes on these two stages, another indicationthat the contactor operation with the TRUEX-NPH solvent was not adequate.
After collecting the rotor samples, we noticed that a thirdphase had formed in the samples collected from stages two through seven. Thisindicated that the solvent capacity had been exceeded. The two organic phaseswere approximately equal in volume, and the darker phase was located at theorganic-aqueous interface. Both organic phases were less dense than theaqueous phase. Further investigation of third-phase formation is discussed inSec. II.G.4.
86
Table 11-17. Checklist for Verification Run 1
ExpectedTimes Actual Times(min:sec) (min:sec) Action Required/Noted
0:00
0:000:15
5:156:007:00
16:00
17:1517:4516:0017:4518:3019:0019:1520:0020:1521:0021:00 - 25:0022:00 - 24:0026:00
47:0048:3050:3053:00
- 15:00- 10:00
0:000:151:002:583:304:206:107:447:0810:3011:5013:0713:3014:1016:0016:0517:1517:5517:5718:0518:3020:4519:0020:0822:3221:00
22:1533:0053:00
54:30
54:4555:15
78:40
Measured feed solution temperaturesSample feed solutions (DF, DS, DX, EF, FF,
GF)Rotors onDS pump on (1st Scrub)Aqueous feed to Stage 8Aqueous feed overflows Stage 8Hood temp 22.5 CAqueous overflows Stage 7DW1 raffinate detected (1st Scrub)Checked DW1 flowrateTurn DX mixer onAqueous overflows Stage 4Checked DW1 flowrateAqueous overflows Stage 3Hood temp 23*CChecked DW1 flowrateEF pump on (2nd Scrub)Aqueous overflows Stage 2FF pump on (1st Strip)EW overflow detected (2nd Scrub)DW effluent detected (Stage 1 Overflow)DX pump on (TRUEX Solvent)GF pump on (2nd Strip)Organic solvent overflows Stage 8FW raffinate detected (1st Strip)GW raffinate detected (2nd Strip)GP raffinate detected (organic solvent)DF pump on (CAW Feed)Check GW for Organic PhaseCheck flowrates on DW, EW, FW, GW, GPBegan sampling effluent streamsAdded 1 L more FF and DS feed tanks
(DS ran out for about 5 s)Valve on GP stream misadjusted (solution
backup in stage 16)End of samplingShut off all pumps and open drain valvesClose drain valvesShut off rotorsSample Stages
87
Because of these problems observed during this test, furtheranalysis of the samples was not completed. New contactor rotors will bedesigned for the TRUEX-NPH solvent. To prevent third-phase formation, methodsfor operating the contactors at elevated temperatures will also beinvestigated.
4. Third-Phase Formation with the TRUEX-NPH Solvent
Because of the formation of the second organic phase in thefirst verification test, a study was conducted to determine the temperature atwhich this phase forms. To conduct this study, six samples of the TRUEX-NPHsolvent were equilibrated with simulated CAW solutions. These samples wereprepared so that each one had a different extractant loading. This wasaccomplished by contacting 3 mL of organic with two different CAW solutions.One was the same solution used in the verification test described above (seeTable 11-16 for its composition), while the second was prepared with this samecomposition, but without Nd3 + and Fe3 +. By varying the number of contactswith the solution containing the Nd3 + and Fe3 +, the extractant loading wasvaried. The five completed contacts are summarized in Table 11-18. Theapproximate solvent loading for Nd3 + is also shown in this table. To estimatethe loading, it was assumed that three molecules of CMPO extract one moleculeof Nd3 +. The solvent loading from other extracted species was not calculated.The phase ratio for these equilibrations was 1:1. For each contact, the twophases were initially equilibrated at 700C for 5 min then vortex mixed 3 timesfor 20 s each.
After the organic solutions had been equilibrated, the sampleswere allowed to stand at the test temperature for approximately 30 min, atwhich time they were examined to see whether a third phase had formed. If athird phase was not detected, the temperature was decreased 100C and themixing and standing steps were repeated. This procedure continued until athird phase was observed.
Results from this test are shown in Table 11-19. For samples 2through 6, a second organic phase formed when the bath temperature was loweredfrom 30 to 20 C. For sample 1, the second organic phase formed when the bathtemperature was lowered from 14.2 to 5CC. For each sample, the heavierorganic phase was the darkest color, implying that the highest concentrationsof TBP and CMPO were in that phase (iron discolors both the aqueous andorganic phases when it is present). Included in this table is the volumeratio between the two organic phases.
For these same samples, the temperature was increased from 200Cin several increments to determine the temperature at which the second organicphase disappears. Results are shown in Table 11-20. At 20 C, the beginningof the test, sample 1 contained only one organic phase. For sample 2, thesecond organic phase disappeared between 25 and 270C, and for the remainingfour samples, the third phase disappeared between 27 and 300C. As before, theheavier organic phase was the darkest color.
Third-phase formation in TRUEX-NPH solvents has been determinedto be extremely dependent on the mix of normal paraffins, with longer chainlength increasing third-phase formation.2 0 The NPH used in preparing thissolvent has Conoco 012-014 having the composition shown in Table 11-21.
88
Table 11-18. Contacts Completed to Equilibrate TRUEX-NPH Solvent
No. Contacts
Sample 2 3 4 5 6
CAW without Nd3 * and Fe3 + 5 4 3 2 1 0
CAW with Nd3 +and Fe3+ 0 1 2 3 4 5
% Loading 0 10 20 30 40 50
'Approximate loading of the CMPO with Nd3 +.
Table 11-19. Temperature Range of Initiation of Third-Phase Formation
Centrifuge Tube Number1 2 3 4 5 6
Temp, *CNo third phase 14.2 30 30 30 30 30Third phase present 5 20 20 20 20 20
Phase Ratioa 1.5 .89 .94 .89 .89 .88
aVolume ratio of the top organic phase to the lower organic phase.
Table 11-20. Temperature of Third Phase Disappearance
Centrifuge Tube Number
1 2 3 4 5 6
Temp Range, 0C N/A 25-27 27-30 27-30 27-30 27-30
89
Table 11-21. Composition of Conoco C1 2 -C 1 4 aUsed in Third-Phase Tests
n-Alkanes Compositionb %
C10 0.27C11 0.63012 18.06013 46.87014 33.42015 0.16016 0.02
99.43
Branched Alkanes Composition,b %
C12 Isomers 0.02013 Isomers 0.27C14 Isomers 0.29015 Isomers 0.0C16 Isomers 0.0
0.58
'VISTA Chemical, Westlake, LA, Batch #M414.bTotal for all isomers.
H. Centrifugal Contactor Development(R. A. Leonard, D. B. Chamberlain, J. C. Hoh, R. A. Benson,J. E. Stangel,* and K. A. Barnthouse**)
As part of the TRUEX Technology Base Development Program, the basicdesign for the Argonne centrifugal contactor is being modified to meet theneeds of specific solvent extraction processes. To evaluate processesinvolving high alpha/beta activity levels (in a glove box) and/or high gammaradiation (in a shielded cell), a contactor was designed and built that can beused where remote handling is required. This new contactor design, built inFY 1987, was evaluated under typical operating conditions in a glove box and ashielded-cell mockup area during this report period. To use the TRUEX-NPHsolvent, the contactor rotor must be modified. To operate the contactor atelevated temperatures, a heating system was designed.
A small contactor for laboratory use, hereafter called a minicontactor,is being designed and built. The first design criterion for the minicontactoris that the liquid volume in each stage be only 1/10 of that in a 4-cmcontactor. If this stage volume could be achieved, the volume of feedsolution needed for flowsheet testing in the minicontactor will be only 1/10
*Co-op student from Georgia Institute of Technology.**Co-op student from University of Cincinnati.
90
that required using a 4-cm contactor. The second design criterion is that theminicontactor must work well at all 0/A flow ratios, just like the 4-cm andlarger contactors do.
In support of this contactor development effort, we are improving ourmethods for measuring and analyzing vibrations. Work being done in this areaincludes the use of (1) the BEAM IV program to model vibrations of thecontactor rotor and (2) proximity probes and Fast Fourier Transform (FFT)devices to measure the amplitude and frequency of contactor rotor vibrations.
1. Con*-.ctors for Remote Handling
In the last report period, the 4-cm centrifugal contactor design wasmodified so that the contactors could be used in a remote facility withmanipulators and a window to provide visual access, and an eight-stage unitwas built. This unit has now been operated in (1) a shielded-cell mockup areawith manipulators and a viewing window and (2) a glove box. In both cases,the design was shown to work as specified. Videotapes were made of theseoperations for future reference. Finally, the hydraulic performance of theunit was tested and altered so that it works as designed.
a. Mockup Tests
Mockup tests were done to establish the remote-handling
capabilities of the remote-handled 4-cm contactor. As shown in Fig. II-15, a
Fig. II-15.
Remote-Handled 4-cm Contactors BeingTested in Mockup of a Shielded-CellFacility
92
The galling problem was corrected by changing from a press fitto a snug fit of the coupling on the motor shaft. The coupling was alsoredesigned to make it stiffer. To this end, the captive nut on themotor/rotor coupling was eliminated by making it an integral part of thecoupling, that is, a coupling nut. For this modified design, the appropriatethreads for a self-tightening connection are right-handed rather than left-handed. With the new coupling nut, the unit works without banging for allliquids, even with tetrachloroethylene, the most dense liquid expected.
Hydraulic tests with two liquid phases showed that both liquidphases flowed as expected. In addition, hydraulic tests were made with oneliquid phase (water). This phase was flowed to a stage, and the flow rate wasincreased until some liquid flowed out the more-dense-phase exit. The flowrate at which this first occurs, called the zero point, was compared with thetheoretical value of 570 10 mL/min for an underflow coefficient, k(U), of1.9 x 104 kg/m3 . The experimental flow rates at the zero point ranged from650 to 880 mL/min for the eight remote-handled rotors, as can be seen inTable 11-22. The reason for this difference between the experimental andcalculated zero point is not known. It was noted that, if no pressure drophad occurred in the underflow (i.e., if k(U) were zero), the calculated zeropoint would be 700 10 mL/min. It was also noted that, for the rotor withthe lowest zero point (rotor 7), the liquid volume in the mixing zone at no-flow conditions was very low, about 25% of the liquid volume that would berequired to fill the space between the stationary vanes below the rotor.
Table 11-22. Experimental Zero Points forthe Rotors of the Remote-Handled 4-cm Contactor
Rotor Number Zero Point, mL/min
1 780 50
2 780 50
3 775 50
4 880 40
5 880 40
6 780 40
7 650 50
8 760 40
2. Rotors for TRUEX-NPH Solvent
In the first verification run with TRUEX-NPH solvent, it wasdiscovered that some of the organic phase followed the aqueous phase into theaqueous effluent from the second strip section. This operation was modeledwith the ROTOR worksheet running in Excel. This ROTOR worksheet is based onthe contactor model given as a FORTRAN program called CCD.2 ' From the model,
92
The galling problem was corrected by changing from a press fitto a snug fit of the coupling on the motor shaft. The coupling was alsoredesigned to make it stiffer. To this end, the captive nut on themotor/rotor coupling was eliminated by making it an integral part of thecoupling, that is, a coupling nut. For this modified design, the appropriatethreads for a self-tightening connection are right-handed rather than left-handed. With the new coupling nut, the unit works without banging for allliquids, even with tetrachloroethylene, the most dense liquid expected.
Hydraulic tests with two liquid phases showed that both liquidphases flowed as expected. In addition, hydraulic tests were made with oneliquid phase (water). This phase was flowed to a stage, and the flow rate wasincreased until some liquid flowed out the more-dense-phase exit. The flowrate at which this first occurs, called the zero point, was compared with thetheoretical value of 570 10 mL/min for an underflow coefficient, k(U), of1 9 x 104 kg/m3. The experimental flow rates at the zero point ranged from660 to 880 mL/min for the eight remote-handled rotors, as can be seen inTable 11-22. The reason for this difference between the experimental andcalculated zero point is not known. It was noted that, if no pressure drophad occurred in the underflow (i.e., if k(U) were zero), the calculated zeropoint would be 700 10 mL/min. It was also noted that, for the rotor withthe lowest zero point (rotor 7), the liquid volume in the mixing zone at no-flow conditions was very low, about 25% of the liquid volume that would berequired to fill the space between the stationary vanes below the rotor.
Table 11-22. Experimental Zero Points forthe Rotors of the Remote-Handled 4-cm Contactor
Rotor Number Zero Point, mL/min
1 780 50
2 780 50
3 775 50
4 880 40
5 880 40
6 780 40
7 650 50
8 760 40
2. Rotors for TRUEX-NPH Solvent
In the first verification run with TRUEX-NPH solvent, it wasdiscovered that some of the organic phase followed the aqueous phase into theaqueous effluent from the second strip section. This operation was modeledwith the ROTOR worksheet running in Excel. This ROTOR worksheet is based onthe contactor model given as a FORTRAN program called CCD.2 From the model,
93
it was found that the poor phase separation was a result of the density of theTRUEX-NPH solvent being fairly close to that of water (only 14% lower). The4-cm rotors were designed to be optimum for PUREX-NPH. Later, it was foundthat this design was also optimum for TRUEX-CC14 and TRUEX-TCE. Using theROTOR worksheet, we designed a new 4-cm rotor that was optimized for use withthe TRUEX-NPH solvent. The results of the design calculations (Rotor II) arecompared with the design of the original 4-cm rotor (Rotor I) in Table 11-23.Basically, the more-dense-phase weir diameter was reduced from 19.05 mm(0.750 in.) for Rotor I to 17.60 mm (0.693 in.) for Rotor II.
Table 11-23. Rotors for the 4-cm Contactor
Maximum TotalWeir Diameter,a Design Range,b Throughput,c
Rotor mm (mils) % mL/min Notes
I 19.05 (750) 18-29 620 d
II 17.60 (693) 10-29 340 e
aFor the more-dense-phase weir.bThis range is the percentage by which the density of the organic phase
is less than that of the aqueous phase. Within this range, the con-tactor will operate satisfactorily up to the maximum total throughput.
cBased on a dimensionless dispersion number of 8 x 10-4. This maximumthroughput applies to a design range for 0/A flow ratios from 0.1to 10.
dRotor to use for 30% TBP in NPH (i.e., the PUREX-NPH solvent). Alsothe rotor to use for TRUEX-TCE (i.e., 0.25M CMPO and 0.75M TBP in TCE).
'Rotor to use for TRUEX-NPH (i.e., 0.20M CMPO and 1.40M TBP in NPH).However, it will also work at these reduced throughputs for thesolvents for which Rotor I was designed.
Rotor II is designed to be operable even when the density of theorganic phase is only 10% lower than that of the aqueous phase. Even forextreme cases where metal ions and acid load the solvent and make itappreciably more dense, Rotor II will still be operable with the TRUEX-NPHsolvent. To gain good operation over a greater range of solvent densities forRotor II, the maximum throughput was reduced to 340 mL/min.
Based on these design calculations, new drawings have been made ofthe rotors for the 4-cm contactors used in "open," "closed," and "remote"operation. These drawings are now being used to make a second set of rotorsfor our 4-cm contactors. These new rotors will be used in future flowsheettests with TRUEX-NPH.
3. Contactor Operation at Elevated Temperatures
In the first verification run, we discovered that, when the TRUEX-NPH solvent was used with the CAW-type waste, a second organic phase formed insome stages. Subsequent batch tests showed that this second organic phasedisappears at slightly elevated temperatures (e.g., 300C). As a result, we
94
are now exploring ways to operate the contactors above this temperature, in
the 40 to 50 C range.
For such an operation, each of the feed liquids would have to bemaintained at the desired test temperature. In addition, it would probably benecessary to heat each contactor stage. Tests in a "closed" four-stage, 4-cmcontactor with 50'C water flowing at 400 mL/min showed temperature losses of0.35 to 0.6 C/stage. Insulating the outer wall of the annular mixing zone andthe interstage lines did not reduce these temperature losses significantly. Atemperature drop of 0.6 C/stage at the above test conditions corresponds to aheat loss of 1.7 W/stage. Various ways to replace this heat are beingreviewed, including sun lamps, band heaters, disk heaters, heating tapes, hotair, and heating mats. Work on developing a method for running the 4-cmcontactor at 50'C is continuing.
4. Minicontactors
A minicontactor is being designed and built for the testing ofsolvent extraction flowsheets. The 4-cm contactor has been used for thispurpose and has done quite we!!; however, it requires about 10 L of feedsolution to run one test. Thr' new minicontactor will be smaller, with muchless liquid holdup, so that oily 1 L of feed solution will be required for atypical flowsheet test. The rotor diameter of the minicontactor will be 2 cm.The rotor internals and the annular mixing zone have been redesigned so thatliquid holdup is less than that of previous 2-cm contactor designs. Inaddition, the rotor has been made longer so that operation will be good at all0/A flow ratios. That the rotor is sufficiently long will be established bybuilding and testing a single-stage minicontactor. In the earlier 2-cmcontactor designs, which had shorter rotors, the units would not work well atcertain low 0/A flow ratios.2 2
a. Motor Selection
The motor chosen for use with the minicontactor is a com-mercially available Bodine Model 710. This motor has a rotational speed of3600 rpm, a torque rating of 0.3 oz-in. (equivalent to 0.80 W at 3600 rpm or1/1000 hp), a shaft diameter of 3/16 in. (0.5 cm), and a motor body width of2 1/2 in. (6.3 cm). Based on the design calculations, the power of this motorshould be more than adequate, as the required power is only 0.16 W.
We also tested the motor to determine if the motor/rotorcoupling can be made stiff enough to prevent rotor banging, that is, todemonstrate that the contactor can operate enough below the first naturalfrequency of the rotating mass (probably 20% or more) to prevent the rotorfrom banging on the housing. To this end, the motor was used to turn a seriesof solid steel rods, each rod having a diameter of 1 in. (2.5 cm). The lengthof this rod was varied from 2.5 in. (6.3 cm) up to 10 in. (25 cm) in 2.5 in.(6.3 cm) increments. These rods, each with a machined shaft at the top, actas stand-ins for the actual rotor. A sketch of the 2.5-in. (6.3-cm) longsolid rotor is shown in Fig. 11-16. A socket set screw is used to hold therotor shaft onto the motor shaft with a gap, typically 1/16 in. (0.16 cm),between the motor face and the end of the rotor shaft. The dimensions for allthe solid rotors and their location on the motor shaft are given inTable 11-24. The location of the dimensions L(2) through L(4) is shown inFig. 11-16, while that for L(1) is shown in Fig. 11-17. Because the rods are
95
7/16-IN. DIA-
L(2)
5/16-IN. ACROSS FLATS A T
L(3)
L(, L(4)i n.
2 1-9/163 5/164 2-1/2
- 1-IN. DIA
Table 11-24.
Fig. 11-16.
Dimensions of 2.5-in. (6.3-cm) LongSolid Rotor
Solid Rotor Dimensions and Location on Motor Shaft
Length, in.Solid Rotor
Code L(1)& L(2)b L(3)c L(4)d Notes
2.5A1 1/16 1 9/16 5/16 2 1/22.5A2 1/16 1 9/16 5/16e 2 1/2 Hollowf2.5A3 5/16 1 9/16 5/16 2 1/2
2.5B 1/8 1 9/16 5/16 2 1/2
5.OA1 1/16 1 9/16 5/168 55.0A2 1/16 1 9/16 5/16 5
5.OB 1/4 1 9/16 5/168 7 1/2
7.5 1/16 1 9/16 5/16 7 1/2
10.0 1/16 1 9/16 5/16 10
aMotor shaft has diameter of 0.1875 in.bRotor shaft diameter is 7/16 in. for all cases.CExcept as noted, shaft diameter is 5/16 in.dOutside diameter is 1 in. Rotor is solid unless otherwise noted.eShaft diameter is 7/16 in. with parallel flats that are 5/16 in.between surfaces.'Solid rotor has a 7/16-in. dia hole drilled 2.25-in. deep intothe bottom end of the rotor.
8Shaft diameter is 7/16 in.
F.
96
I
SOLIDROTOR
L(1)
HOUSING
710 DODINE MOTOR
IG
I
SOCKET SET SCREW6-32 UNC X 1/8IN LON
ROTOR SHAFt
L(M,
S1/16
-. 1-1/4 IN. DIA.*-
Fig. 11-17. Motor/Rotor Assembly Mounted inTubular Housing with Support Arm
solid, they have a mass greater than an equivalent contactor rotor of the samedimensions. This means that the natural frequency measured for each solid rodwill be conservative, i.e., the natural frequency will be lower than that forthe corresponding contactor rotor. Thus, an actual contactor rotor will runeven better than the solid rotors.
The solid rotor was run in a steel housing with an insidediameter of 1.25 in. (3.17 cm), as shown in Fig. 11-17. This allowed us todetect any major vibrations easily as the rotor would bang on the housing. Italso allowed us to determine the effect of water in the annular mixing zone onrotor vibrations. This is one feature that cannot be modeled with thesophisticated BEAM IV program for vibration measurement and analysis(Sec. 11.1.5). The results of these tests are given in Table 11-25. Itturned out that having the motor/rotor/housing system clamped was a moresevere test than having it unclamped. The actual rotor will be mounted in ahousing that can be considered to be clamped because of its weight. Itappears that water in the annular mixing zone between the rotor and the rotorhousing will help to damp out vibrations in the actual centrifugal contactor,which should behave like the clamped rotor setup here. Only for rotor 2.5A2was the design such that it did not bang for any of the tests. Because rotor2.5A2 is hollow in the center and is only 2.5-in. (6.3-cm) long, it is also
k ARM FOR CLAMPING DOWNTHE ROTOR HOUSING WITHTHE MOTOR/ROTORASSEMBLY ATTACHED
III
rms-o" IIi
I
Table 11-25. Mechanical Performance of Solid Rotors
Startup
OperatingConditionsa
2.5A1
BangMotorb (Y/N)
UAUwCACw
UACACW
UAUwCACW
UAUwCA
UAUwCACW
UAUwCACW
UAUWCACW
AAAA
BBB
BBBB
AAA
AAAA
BBBB
AAAA
Running
Vibration,C in./sBang
(Y/N)
NNN
d
NYyf
NNNN
NNNN
NNN
NNNN
NNN
NNNN
YYYY
NNNN
NNYN
N1'NYN
NNYN
Pos. 1 Pos. 2
0.70
0.86
1.81
0.97
0.25
0.0670.058
0.028
0.160.063
0.02
0.0960.066
0.070
0.78
0.88
1.82
0.97
0.25
0.0770.061
0.006
0.170.069
0.02
0.0960.066
0.005
Shutdown
Bang
(Y/N)
NNNN
NNN
NNNN
NNN
NNNN
YNNN
NNYN
Time for MaxVibration,
s
None"if"
"n
"n
"n
5None
"n
"n
43.5
None"
2.52
2
44
None"
Notes
SolidRotor
Code
NNY
2.5A1
2.5A2
2.5B
5.OA1
5.0A2
5. OB
eee
gggg
Co
Table 11-25. Mechanical Performance of Solid Rotors (Cont'd)
Startup Running Shutdown
Solid Vibration,C in./s Time for Max
Rotor Operating Bang Bang Bang Vibration,
Code Conditionsa Motorb (Y/N) (Y/N) Pos. 1 Pos. 2 (Y/N) s Notes
7.5 UA B Y Nh 0.09 0.14 Y 9UW B Y Nh 0.04 0.04 Y 5CA B N Y -- -- N None
CW B Y Y -- -- N
10.0 UA B Y N 0.08 0.08 Y 14UW B Y Y -- -- Y
CA B Y Y -- -- N --
CW B Y Y -- -- N --
'U indicates that motor/rotor/housing system was not clamped down; C indicates that it was. A indicates
test was made in air. If the system was unclamped when the test was made i1' air, it was held by hand.
W indicates test was made with the solid rotor in water. If the system was unclamped when the test was
made in water, it was resting on top of a graduated cylinder.
bMotor A had serial number NH4183. It was used for the initial tests. When we were unable to repeat the
UW test for rotor 2.5A1 without it banging, we noted that the motor shaft was slightly bent and used
Motor B with serial number NH4191 for all future tests.
cThese two columns give Vibropac I, model 675 (Balance Technology, Ann Arbor, MI) measurements.
Position 1 is near top of rotor housing with Vibropac sensor opposite the clamping arm on the housing.
Position 2 is near top of rotor housing with Vibropac sensor perpendicular to the clamping arm. The
sensor is on the right side when the clamping arm is to the back.
dNot run initially. On repeat test where UW test banged, we also got banging here. In addition, water
was sucked up through the rotor housing to the motor face.
eThis solid rotor is slightly loose on motor B as it was initially fit to the shaft of motor A.
'Water was sucked up through the rotor housing to motor face.
gHollow rotor. In total mass it is closest to actual contactor rotor filled with liquid. However, the
mass of this hollow rotor is still greater than that of the actual contactor rotor.
hInitial wobbling caused contact between rotor and housing; however, once the vibration was damped by
grasping the housing by hand, the assembly ran smoothly.
Co00:
the rotor with the least mass. An actual contactor rotor that is 0.875 in.(2 cm) in diameter and 2.5-in. (6.3-cm) long will have even less mass, as theoutside diameter is slightly smaller and some of the metal volume still leftwill be replaced by a less dense liquid. This lower mass will result in thecontactor rotor operating even further below its first critical speed. Thus,the Bodine 710 motor can be used if the 2.5-in. (6.3-cm) long rotor is longenough to solve the phase inversion problem seen with the earlier 0.875 in.(2-cm) contactors.
From this brute force approach, we have gained enough infor-mation on mechanical vibrations to design and build a single-stage contactorthat will have a satisfactory low level of vibrations. The rotor will be0.875 in. (2-cm) in diameter and 2.5-in. (6.3-cm) long. If the rotor needs tobe much longer than 2.5 in. (6.3 cm), we would need to find a different motorthat would prevent banging of the rotor on the contactor housing. InSec. 11.1.5, this brute force method is confirmed by a sophisticated approachwhich combines the BEAM IV program with the measurements from the Zonic RealTime Spectrum Analyzer.
b. Rotor and Housing Design
The design of the rotor for the minicontactor has beencompleted and is now being fabricated in the CMT shop. In this new 2-cm rotordesign, the rotor was made 60% longer than earlier 2-cm rotors to prevent thedispersion in the mixing zone from backing up and overflowing into the lowercollector ring. The new rotor will have to be tested to determine if it islong enough to prevent this problem. The annular gap is small, 1/16 in.(0.16 cm), to keep the extraction efficiency in the mixing zone above 98%.Based on the earlier 2-cm rotor tests, a small annular gap limits the range of0/A flow ratios, where either phase can be the continuous phase. Thus, itshould help to solve the phase inversion problem. It will also reduce theliquid volume in the mixing zone and so help to get the total liquid volume inthe stage to 10 mL. To further reduce the liquid volume in the stage, the twoweir diameters were increased so that the volume in the separating zone insidethe rotor would be decreased. Then, because the rotor weirs are larger, therotor inlet diameter was also increased by 50%. Another place to reduceliquid volume is to eliminate most of the dead volume in the separating zone,that is, the liquid that is in a region where the radius is greater than theinner radius of the underflow. Only enough of this region is left to get theliquid to the underflow holes. For ease of fabrication, the underflow holeshave been replaced by four underflow slots, and the diverter disk in theseparating zone was eliminated.
One of the key criterion for rotor design was to reduce thecontactor stage volume for the 2-cm contactor to only 1/10th that of the 4-cmcontactor. In the final design for the new 2-cm rotor, the liquid volume in a2-cm (minicontactor) stage will be 8.9 mL, compared with 100 mL for a typical4-cm stage. The other key design criterion, good operation over the entirerange of 0/A flow ratios, can only be established in actual single-stagetests. As mentioned previously, the rotor was made longer and the annular gapwas made smaller to meet this second design criterion.
100
Two contactor housings are now being designed. The one housingwill be transparent (acrylic) so that we can evaluate mixing zone operationand determine if its increased length is sufficient to give good contactoroperation at all 0/A flow ratios. The other housing will be stainless steelso that we can operate the contactor with TCE and be sure that the vibrationsare at an acceptably low level for this worst case, i.e., where the liquid inthe rotor has the highest density possible.
5. Vibration Measurement and Analysis
Techniques for vibration measurement and analysis are now beingincorporated into our contactor development effort to improve contactor designand reduce development costs.
a. Analysis
The BEAM IV computer program, developed at Virginia PolytechnicInstitute and State University using BASIC and now installed on the CMT VAX,is being used to model vibrations in the system consisting of the contactorrotor combined with the motor rotor. It can be run from a Macintosh computerusing VersaTerm, software that does a good emulation of a Tektronix terminal.The BEAM IV program allows us to maximize the mechanical information that weextract from our exploratory tests with small motors for use with theminicontactor. In particular, we are usin. the program to determine, inconjunction with experimental measurements, the spring constants for the motorbearings. With these spring constants, we can then calculate the criticalspeeds of the spinning motor/rotor system before we actually build the rotor.The overall result is a great savings in time and cost for contactordevelopment.
Using the results from a demonstration of the Zonic Real TimeSpectrum Analyzer (RTSA), as well as additional tests made by T. M. Mulcahy[ANL Materials and Components Technology (MCT) Division], we have developed aBEAM IV model for the Bodine Model 710 motor called MINI.002. To this basicmotor model we added additional sections so that we could model the solidrotors described above. The results of these BEAM IV models for the solidrotors in terms of the first and second critical speed (first and secondnatural frequencies) are shown in Table 11-26. All models assume (1) a1/16-in. (0.16-cm) gap between the end of the rotor shaft and the motor faceand (2) a 5/16-in. (0.8-cm) diameter for the 5/16-in. (0.8-cm) length of rotorshaft L(3) just before the main body of the solid rotor. Other dimensions aregiven in Table 11-24. Note that only those rotors that are 2.5-in. (6.3-cm)long have first natural frequencies that are higher than the motor speed of60 Hz (3600 rpm). Note also that the second critical speed is higher than thefirst critical speed by a factor of 30.
The BEAM IV results given in Table 11-26 are for the rotatingsystem in the forward whirl mode. This model gave slightly different valuesfor the first natural frequency fn in the static mode and the backward whirlmode. For the static mode, fn was found to be slightly less (3 to 17%) thanthat for the forward whirl. For the backward whirl mode, fn was found to beslightly less (3 to 13%) than that for the static whirl. These results,
101
Table 11-26. Critical Speeds for Solid Rotors of the Minicontactor Design
First Critical SecondSolid Rotor Speed, Hz Critical
Model Rotor Length, Hollow SpeedName Code in. (Y/N) Meas'db Calc'd (Calc'd), Hz Notes
MINI.003 2.5A2 2.5 Yes - 86.1 2097 - C
MINI.004 2.5A1 2.5 No - 79.2 2458MINI.005 5.0A2 5.0 No 37.5 2.5 40.4 1320MINI.006 7.5 7.5 No r6.0 2.5 25.7 829MINI.007 10.0 10.0 No 17.5 2.5 18.2 603MINI.003 2.5A2 2.5 Yes - 66.3 1850 d
aThese models correspond to files on the CMT VAX. Use "BEAMS" to access thesefiles. The "S" in BEAMS indicates that single precision was used in thethe calculations.
bUsed strobe to measure this critical speed.cAssumes a diameter of 5/16 in. for the section whose length is given by L(3).dIncreased the gap between the motor/rotor coupling and the motor face by0.25 in., from 1/16 in. to 5/16 in.
discussed with L. D. Mitchell, Virginia Polytechnic Institute and StateUniversity, are about what we should expect. They are caused by the center ofgravity of the rotating rotor being off-center as it vibrates. Thus, thesmallest differences are for the short [2.5-in. (6.3-cm) length] solid rotorwhich is hollow. We will focus on the forward whirl mode for most cases. Thestatic mode is important if a contactor rotor is not rotating. The backwardwhirl is important only if the rotor is in a confined space and starts hittingthe housing. As this is possible in our situation, we plan to design units sothat we are also 10 to 20% from a critical speed for a backward whirl.
b. Measurements
To make the vibrational measurements needed to use the BEAM IVprogram, we purchased (1) a Zonic Real Time Spectrum Analyzer (RTSA), (2) animpulse hammer kit to use with the RTSA, and (3) a Kaman proximity probe thatcan measure motor shaft movement when the contactor is operating.
Using the Zonic RTSA (also called the Zonic 6088 multichannelsignal processor and built by Zonic Corp., Milford, Ohio) and assisted by ArtRich of Zonic Corp., we measured the first three resonant peaks for two motorsof interest, the Bodine 710 motor and the Bodine NSI-13 with serial numbersM0052xxx. These motors are used with the 2-cm and 4-cm contractors,respectively. In addition, T. M. Mulcahy (MCT) helped us use (1) anaccelerometer with an oscilloscope and (2) a Hewlett-Packard (HP) multichannelanalyzer to obtain some additional vibration readings. The results, shown inTable 11-27, give the first three resonance peaks for both motors. For boththe Zonic RTSA and the HP multichannel analyzer, an accelerometer on the motorshaft was connected to the first analyzer channel, and a force,-reading impulsehammer, used to strike the motor shaft, was connected to the second analyzerchannel.
102
Table 11-27. Vibration Tests on Two Types of Contactor Motors
Bodine Resonance Peak, HzMotorModel Technique First Second Third
710a,b,c Accelerometer -- --- 5200
withOscilliscope
710b,c Zonic RTSA -- --- 5260, 5380
710d'* Zonic RTSA 2430 3888 ----
710b,c HP Multichannel (2300) 3472 5264, 5360Analyzer
NSI-13df Zonic RTSA 2730 3000 31352745
'This Bodine motor is the type that we would like to use with theminicontactcr.
bSerial number is 710NH4185.cAccelerometer was cemented to motor shaft.dBetswax was used to hold the accelerometer to the motor shaft.eSerial number is 710NH4154.IThis Bodine motor, serial number M0052027, is the type that we have been
using for the 4-cm contactors.
The first three resonance peaks for the Bodine 710 motor wereused, in conjunction with the BEAM IV program, to determine the bearingstiffness for this motor. This bearing stiffness had a value of2.4 x 105 lbf/in. (4.2 x 107 N/m). The BEAM IV modeling also showed that thethird resonance peak, which was actually two peaks, represented the ringing ofthe motor shaft. This double peak probably derives from the flat on the motorshaft, since the area moment of inertia for the shaft differs depending onwhether one determines it parallel or perpendicular to the flat. When thecontactor rotor is attached to the motor shaft, BEAM IV modeling shows thatthis third resonance peak drops sharply, becoming the first critical speedlisted in Table 11-26. The first resonance peak in Table 11-27 then becomesthe second critical speed listed in Table 11-26. This speed drops only slowlywith increasing rotor mass and length.
With the setup shown in Fig. 11-17, the first critical speedwas measured for the three solid rotors listed in Table 11-24 (rotors 5.0A2,7.5, and 10.0) which have values below 60 Hz. This measurement was made byusing a stroboscope on the marked bottom face of the rotor. At a fixed stroberate, the motor was shut off, and the bottom face of the solid rotor waswatched as the motor/rotor system coasted to a stop. It was noted whether themaximum rotor vibration occurred above or below the speed at which the strobewas flashing. In this way, the first critical speed of the three longestrotors was determined to within 15% or better. The experimental results,given in Table 11-26, show reasonable agreement with the calculated values.
103
This agreement serves as a check of the value for the stiffness of the motorbearings found using BEAM IV with the results of the Zonic RTSA tests on theBodine 710 motor when it was stationary.
Note that the first critical speed for the solid 2.5-in.(6.3-cm) long rotor (2.5A1) is 32% above the motor speed. Since a firstcritical speed that is 20% above the operating speed should normally besufficient for good operation based on American Petroleum Institute (API)standards, the 2.5A1 rotor should be vibration free. However, the actual testresults in Table 11-25 are mixed. As mentioned above, this may be the effectof the backward whirl, which has a first critical speed that is 6 to 30% lowerthan that for forward whirl. The bent motor shaft could also be a factor. Atany rate, the hollow 2.5-in. (6.3-cm) long rotor, 2.5A2, which has a firstcritical speed that is more than 40% above the motor speed, operates with asatisfactory low level of vibration and does not bang the rotor housing.
I. Radiclysis and Hydrolysis of TRUEX-NPH Solvent(N. Simonzadeh and L. E. Trevorrow)
1. Introduction
During treatment of PUREX raffinates from the reprocessing ofirradiated fuel, a fraction of the TRUEX-NPH solvent will be converted,through radiolysis and hydrolysis, to other chemical species. The purpose ofthis study is to (1) derive parameters that can be used in the Generic TRUEXModel to express the effects of radiolysis and hydrolysis on the extractionproperties of the solvent and (2) develop procedures for removing the productsof radiolysis and hydrolysis from TRUEX-NPH solvent.
Studies of the hydrolysis and radiolysis of various TRUEX solventswere carried out earlier.2 -25 The present study will be concernedspecifically with hydrolysis and radiolysis of the TRU3X-NPH solvent (0.2MCMPO and 1.4M TBP in Conoco C12-C14).
2. Procedure
Samples of TRUEX-NPH solvent degraded by radiolysis in contact witheither 0.25 HNO3 or 6.OM HNO3 were prepared by exposure to a 6 0Co source.
A set of 400-mL thermostated beakers (500% filtered with ref luxcondensers was arranged in a hot cell containing a 6 Co source, so that eachvessel was exposed to (2.5 0.25) x 105 rad/h. The beakers contained equalvolumes (100 mL of each phase) of the TRUEX-NPH solvent and acid, whict. weremixed vigorously. Samples of the radiolytically degraded solvent were used inmeasurements of the distributior of 24 1 Am between TRUEX-NPH solvent and either0.01M, 0.05M, or 2.OM HNO3 . Some of the degraded TRUEX-NPH samples were alsoscrubbed with a 0.25M sodium carbonate solution before measurement of Dmvalues for the three acid concentrations.
104
Total dose received by the solvent was obtained from the exposuretime and the power density, which was calculated according to the algorithm
2.5 x 10 rad/h] x
85.9 erg
E(TRUEX) rad 3
'1xl10- 7 Wserg x 3600 s
(II-30)
(0.854 g/mL ) x (103 mL/L) = 0.5
This calculation includes the ratio of electron concentration of the TRUEX-NPHto that of water (i.e., 100 erg/g(H20)erad x (477Me/555Me) =85.9 erg/g(TRUEX)*rad). The density of the TRUEX-NPH solvent is 0.854 g/mL.This method of calculating total dose is a standard one that takes intoaccount the electron concentration difference between the TRUEX solvent andH20.
3. Results
The DAm values for degraded and non-degraded solvent are showngraphically in Figs. 11-18 through 11-23. These data indicate that carbonatescrubbing of the degraded solvent generally results in lower DAm values for0.O1M and 0.05M HN03 (i.e., it removes acidic products), whereas it improvesamericium extraction (high DA. values) at 2.OM 1N03 . The carbonate scrub,however, seems to be only partially effective in restoring solvent behavior in2 4 1Am extraction and stripping; a search will be made for better scrubreagents.
0
0
100
10
1
.1
0 100Wh/L
0DAm not ScrubbedDAm Scrubbed
200
Fig. 11-18. Extraction of Am from 0.O1M HNO3 by TRUEX-NPH.Solvent was previously irradiated while incontact with 6.OM HNO3 .
iiiilllllllllllllllllIllllilllllliiiiiilllIllilljlllllllllllllllilillllll1111I
C
_
-v-
DAm not ScrubbedDAm Scrubbed
Fig. 11-19.
0
N4-1
E0
100
10
1
0
Fig. 11-20.
Extraction of Am from 0.05M 1HN03 by TRUEX-NPH.Solvent was previously irradiated while incontact with 6.OM HNO3 .
100W/L
DAm not ScrubbedDAm Scrubbed
200
Extraction of Am from 2.OM HN03 by TRUEX-NPH.Solvent was previously irradiated while incontact with 6.OM HNO3 .
105
10,*10
0
0 .11
0
.010 100
Wh L200
I
106
20 40 60WhL
DAm not ScrubbedDAm Scrubbed
80 100 120
Fig. 11-21.
1
4
4
4
.1 -
.01
Extraction of Am from O.O1M HNO3 by TRUEX-NPH.Solvent was previously irradiated while incontact with O.25M HNO3.
0 20 40 60Wh/L
Fig. 11-22.
80 100
DAm not ScrubbedDAm Scrubbed
120
Extraction of Am from O.05M HNO3 by TRUEX-NPH.Solvent was previously irradiated while incontact with O.25M HN03 .
10
0
O-
E
.1 1
.01
.001 a
0
0
LO)0
E0
-. - ----.-.
I- U
I Avol4k
it
4
107
100
c>. 0 DAm not ScrubbedDAM Scrubbed
E
0
0 20 40 60 80 100 120Wh/L
Fig. 11-23. Extraction of Am from 2.OM HNO3 by TRUEX-NPH.Solvent was previously irradiated while incontact with 0.25M HNO3 .
J. PUREX-TRUEX Processing of Chloride Salt Residues(L. Reichley-Yinger, D. R. Fredrickson, R. A. Leonard, andG. F. Vandegrift)
1. Introduction
Los Alamos National Laboratory (LANL) is testing PUREX-TRUEXflowsheets for extraction cf plutonium and other TRU elements from acidicchloride-containing residues that are produced during plutonium-metalproduction. We were funded in 1987 to support the LANL efforts with flowsheetdevelopment activities encompassing laboratory and computer simulationstudies. Flowsheets must be designed to (1) remove Pu and Am from the feed tothe extent necessary to make the solution a discardable, if not a nonTRU,waste, (2) separate a purified Pu stream that is acceptable for Pu-metalproduction, and (3) accomplish these objectives in 16 stages of centrifugalcontactor equipment contained in a glove box.
2. Plutonium Model Development
A model that predicts the Pu(IV) distribution ratio between TBP inTCE solutions and acidic chloride media was completed. The model incorporatesaqueous-phase complexation equilibria as well as extraction equilibria tocalculate the distribution ratio for Pu(IV). Formation constants (p values)for Pu(IV)-chloride complexes, which are based on activities, were evaluatedfrom literature and our own experimental data. These P values were, in turn,used to determine equilibrium constants for the three extraction reactionsbetween plutonium-chloride species and TBP. The model, which has six con-stants, three P values, and three extraction equilibrium constants, was foundto predict the Pu(IV) distribution ratio over four orders of magnitude, from10~ to 102 of hydrochloric acid activities.
108
Formation constants for PuCln(H 2 0)gn (4-n)+, iPn, from the reaction
Pu(H2 0)g4 + + nCl - PuCln(H 2 0) 9-_C 4 -n)+ + nH2 O (II-3)
were calculated from thenoyltrifluoroacetone (HTTA) extraction data using thefollowing:
{H20}"
{H20}8]
Do
D
{C1-}
{H20}
{01~}2
P2 {H20}2
The calculation required that the activities of aqueous phase H+, Cl-, and H2 0be known, as well as the distribution ratios for Pu(IV) between HTTA inbenzene and HC1-HC104 solutions. The single-ion activity coefficients of H'(species i=1) and 01~ (species j=2) were calculated using Bromley's method:26
-Az (1)1/2log7 -
1 + 11/2
Az (1)1/2log 7 =
1 + 11/2
+ F 1 (II-33a)
+ F2 (II-33b)
where A = 0. 11 kg" 2 mol-1/2 (the Debye-Huckel constant for 250C), z1are the ionic charges of H+ and C1~, and I is the molal ionic strengthsolution. The parameters F1 and F2 are defined for H+ (i = 1) and C-as follows:
and z2of the(j = 2)
F1 = B1 2m2 + B14m4
F2 = Biami
where m is the molarity of each species and j = 4 refers to C104~.
Eqs. 11-34 and 35, B1 j is defined as follows:
Bij = (0.06 + 0.6B1 3)(zizj)/(1 + 1.5I)2 + Bid
(II-34)
(II-35)
For
(II-36)
with B12 = 0.1433 and B1 4 = 0.1639. Water activities were calculated using2 7
= X1Y2log{H20}0 12 + X1Y4 log{Ha0}*14
(II-32)
log{H20}mix (II-37)
109
where Xi = Ii/I, and Yj = Ij/Ia, with I, and Ia being the molal ionic strengthof the cations and anions in solution, respectively. The water activities ofthe pure acid solutions, {H20}
01 1, were calculated from
log{H20}012 = 0.0062 - 0.02041 - 0.001712
log{H2 0}*14 = 0.0021 - 0.01281 - 0.003712
(II-38)
(II-39)
These equations were obtained from curve-fitting of literature data.28applying Eq. 11-32 to the literature data,2 9 we obtained Pl = 0.354 and
p: 0.177.
On
The molal concentrations of aqueous solutions were calculated fromt.hA molar concentrations as follows:
mi = ci/(psoin - Ei 10~ 3 (MW) ici) (II-40)
where m is the molal concentration in (moles of solute i) kilograms ofsolvent, c is the molar concentration in (moles of solute i) liters ofsolvent, Psoin is the density of the solution, and MW is the molecular weightof species i. For HCl-HC104 solutions, the density of solution isapproximated by
PsoIn = 1.31 - 0.0606cHC1 (11-41)
Equation 11-41 was obtained from the measured solution densities shown inTable 11-28. The concentrations of the HC104 and HC1 stock solutions were8.690 and 10.64M, respectively. The solution densities for acidic chloridesolutions were measured and reported previously.'8
Table 11-28. Densitiesat 25 C
of HC1-HC104 Solutions
[HCl], [HC104], Solution Density,M M g/mL
0.491 3.664 1.2789 0.0001
0.887 3.141 1.2505 0.0001
1.965 2.094 1.1882 0.0005
2.947 1.046 1.1283 0.0005
3.439 0.523 1.0996 0.0005
4.00 - 1.066a
aInterpolated from literature data.3 0
110
Values for /33 and /34 were obtained from experimental data by usingPi and /32 and the extraction equilibria K1 , K2 , and K3 , which are described asfollows:
{C1} 4 [TBP]free
Dpu{H20}7
{HCl}K1 + K2
{H20}
= 1 + 3H +}
{H20}
{01-}2
P2 {H 2 0} 2
{HCl}2
+ K3{H20}2 J
{C1-}3
+ r3 {H20}3(II-42)
From the data for which 15 < {01-} < 50, a value of 0.016 was calculated for/33. A /34 value of 6.8 x 10~4 was calculated from the data for which{Cl-} > 90.
Extraction constants K1 , K2, and K3 were obtained from the experi-mental data with {C1-} < 9 by using
{Cl-}4 [TBP]Iree
{H20}7K1
{Cl-}1 + /30
{H30}
{HCl}+ K2
{H20}
{01}2+ /32
{H120}2
{HCl}2+ K3
{H20}2
{11-2}
{H20}3
The extraction constant, Km+i (m = 0,1,2) is defined by the reaction
Pu(H 2 0)g4+ + mH+ + pCl- + qTBP <
(II-44)Pu(H20)g9_(p+q) Cl 4"qTBPemHC1 + (p + q) H20
Values of 3.4 x 10-5, 7.6 x 10-7, and 1.8 x 10-8 were determined for K1, K2 ,and K3 , respectively. Figure 11-24, a log-log plot of the measured andcalculated Pu(IV) distribution ratios as functions of the {C1}, shows howwell Eq. 11-43 predicts the distribution ratios for the HC1-only solutions.Calculated and measured Dpu values are presented in Table 11-29 for acidicsalt solutions containing NaCl, KCl, and MgCl2 . Further calculations areneeded to determine Dpu values for acidic salt solutions containing CaCl2 .
K. Production and Separation of 9 9 Mo from LEUJ. D. Kwok, W. E. Streets, and G. F. Vandegrift)
1. Introduction
Generators produce 9 9mTc (t1/ 2 = 6.02 h) for medical purposes from
9 9Mo (t1/ 2 = 66.0 h) that is produced in nuclear reactors as a fission product
] (II-43)
Fig. 11-24.
-1
S-- Caclted
2
Table 11-29. Calculated and Measured Pu(IV) Distribution Ratios between25% TBP in TCE and Acidic Chloride Solutions at 25*C
Conc., M Distribution Ratios
[HCl]a [NaCl] [KC1] [MgCl2] [CaCl2] {H}b {Cl}b Calc. Meas.
1.86 1.2 1.25 0 0 4.5 5.5 1.92E-02 3.83E-02
1.62 0.70 0.74 1.12 0 5.33 11.4 2.34E-01 1.62E+00
5.61 0.19 0.20 1.30 0 33.6 35.1 4.96E+02 4.40E+02
5.22 0.10 0.10 0.87 0 38.6 47.5 1.08E+03 1.11E+03
1.26 0.30 0.31 2.6 0 8.07 32 2.66E+01 2.60E+02
5.14 0 0 0 1.77 77.8 94 7.80E+02 4.04E+03
aTitrated values.bCalculated using Bromley's model.2 9
of 2 3 5 U or by the (n, 7) reaction of 9 8Mo (23.7% of natural molybdenum). Oureffort is concerned with producing fission-product 9 9Mo from low-enricheduranium (LEU, <20% 23 5 U). Presently, most of the world's supply of fission-product 9Mo is produced in targets of high-enriched uranium (HEU, '93% 23 5U).
The United States is considering prohibiting the export and internal commer-cial use of HEU because of its potential for use in an atomic bomb. For thepast eight years, the Reduced Enrichment Research Test Reactor (RERTR) program
4)
111
o.
0or0
-2
-40 1
log(CI-)
1 l
T
T
Calculated and Measured Pu(IV)Distribution Ratios vs. HClActivity at 25 C
112
office has been developing reactor and reactor-fuel designs to accommodate theuse of low enriched flux. The next goal is to reduce the 2 35 U content oftargets used to produce 9 9Mo from HEU to LEU.
During this six-month period, we have investigated the sorption anddesorption conditions of molybdenum on alumina columns. An irradiation ofU3 Si2 (LEU) has been performed in the ANL Janus reactor. This target has beendissolved, and the 99 Mo was separated from other components on an aluminacolumn. Gamma spectrometry has been used to measure the activities of fissionproducts and neptunium in the various fractions. Results of these separationsare presented.
2. Sorption and Desorption of Molybdenum on Alumina Columns
a. Experimental
Alumina column experiments were run using a setup incorporatingjacketed alumina columns, a peristaltic pump, automatic sample collector, andtracer "9 Mo for analyses. As a source of 9 Mo, a 2.7 Ci 99mTc generator waspurchased from Du Pont NEN along with its circular "Molycoddle" leadshielding.
Parameters that were varied include column temperature,concentration of the nitric acid loading solution, flow rates, and washsolutions. All runs were eluted with concentrated NH4OH. These experimentalparameters are summarized in Table 11-30.
An automatic sample collector was used to collect a discrete1-mL aliquot of the eluate from the wash solutions and the concentrated NH4OHeluant into sample vials. The vials were capped, and the gamma activity ofthe 9 9 Mo peak at 181 keV was determined for each vial using a NaI detector.After gamma counting, the pH of each vial was measured. Figures 11-25 through11-29 show the gamma activity and pH of each aliquot for runs 1 through 5,respectively.
Table 11-30. Experimental Parameters for 9 9 Mo Runs
Conc. of HNO3Temp., Flow Rate, Loading Solution
Run No. C mL/min [50 mL], M Wash Solutions
1 25 1.2 0.5 25 mL 0.5M HNO3 + 50 mL H20
2 25 0.67 0.5 25 mL 0.5M HNO 3 + 50 mL H20
3' 25 1.0 1.0 25 mL 0.01M NH4OH
4' 25 1.2 1.0 25 mL 0.01M NH40H
5' 50b 1.2 1.0 25 mL 0.01M NH4OH
aIntroduction of eluant via syringe to eliminate head space above column.bTemperature had to be lowered to 25*C during elution with concentrated NH4 0H
to prevent boiling of the eluant.
113
2e.3 T14
Fig. 11-25.
Elution of 9 9Mo from Alumina Run 1.(Double peak may be due to poor mixingof eluant in head space above column)
-
-12
-10
-8
-6
4
-2
o.o t:~r .. U20 40
Volume of Column Effluent,mL
- 12
-11 Fig. 11-26.
Elution of 9 9Mo from Alumina Run 2
-10 =
.9
-8
5
Volume of Column Effluent,mL
0-cpm
pH
z Mo-99 cpmt.-fr
1e+5
8e+4
6e+4-
4e+4 -
0 5 10
1 A
--
IA+S 17
ue~ o . . . ,n o _ .111-
-
114
6e+4-
Fig. 11-27.
Elution of 9 9Mo from Alumina Run 3b
U
. A
10 20 30 40 50 60
Volume of Column Effluent,mL
14
- cpm10-22
6e+4
4e+4 1
2e4 8
0 20 40 60
Fig. 11-28.
Elution of 9 9Mo from Alumina Run 4
0.
80
Volume of Column Effluent,mL
r
0.U
Rai 44
5e+4 - Mo-99 cpm
4e+4
1a
3e.-4
-8
2e+4
61e+4
2
0
?I
0
115
12
69+4
11
Fig. 11-29.4e+4 10
Elution of 99 Mo from Alumina Run 5 + "0-23 9
-9
2e+4-
8
Oe+0 -7
0 10 20 30 40
Volume of Column Effluent,mL
b. Results and Discussion
Figures 11-25 through 11-29 show a strong correlation betweenthe pH of the sample and the amount of 9 9 Mo that exits the column. Thismolybdenum is evidently desorbed very rapidly when the pH of the eluant rises.The most disturbing point here is that the a Mo (>80%) appears to be elutingwith the 0.01M NH4OH wash in runs 3-5. Further experiments dealt wibh thiseffort (Sec. II.K.2.c).
Problems with polymer formation, encountered during earlierbulk studies,3 1 should not be involved here because of the use of fission-product 99Mo. This means a maximum Mo concentration of 10-8 M, well below the10-3 M limit for polymer formation in solution.3 3 However, there still remainsthe possibility of polymer formation on the column. Each plot shows anadditional small peak appearing when the concentrated NH40H (pH > 12) comesthrough the column.
The resolution of the large peaks, as measured by the fullwidth at half maximum (FWHM in Table 11-31), remains fairly constant for allthese runs with the exception of Run 2, probably due to its lower flow rate.As shown with the bulk studies, kinetics plays an important role in theadsorption of molybdenum. Whether the better resolution is real and due tothe adsorption or desorption step cannot be determined from this information.
To determine the amount of 9 9 Mo left on the column afteradsorption, the effluent for loading and washing was evaporated and counted(Table 11-31). This plus the total counts for the elution samples wasinvariably over 100% of the activity originally loaded onto the column, adiscrepancy that makes it impossible to determine the 9 9 Mo remaining on the
8e+4 .-- r I I
116
Table 11-31. Experimental Results for Elution of Mo-99 from Alumina Column
Activity, cpm
Total Mo-99 Total Obs.Run FWHM, Mo-99 Mo-99 in Total Calc.,No. mL Eluted Effluent Observed Calculated %
1 - 854600 1000 855600 593600 144
2 2.3 796300 200 796500 763000 104
3 4.5 453300 140 453400 443000 102
4 5.0 36700 4770 541400 421300 129
5 4.0 200000 - - 192000 104
column. It appears that less than 1% elutes during loading and washing. The9 9Mo remaining on the column may be important and has been reported to be asmuch as 10% for actual irradiated 2 3 5U-silicide targets.3 3 The explanationfor the discrepancy between e pected and observed 9 Mo is probably due to thecounting techniques.
c. Further Investigations
Several possibilities for the apparently early elution of Mowith the 0.O1M NH4OH were postulated, including incomplete conversion of Mo toMo(VI) and neutralization of the alumina column by the wash solution.
A change in counting conditions for the determination of thegamma activity was also investigated because of the apparent over-recovery of9 9Mo in the previous experiments.
(1) Experimental Conditions
To ensure the presence of Mo(VI), approximately 100 L of30% H202 was added to the 10 mL of 9 9Mo in concentrated ammonium hydroxideeluted from the Tc generator. After about 30 min, bubbles began forming andcoming off.
The alumina (3 g) was washed with water and conditionedwith 50 mL of 0.5M HN03 . After 50 mL of 0.5M HNO3 was spiked with 99Mo, itwas loaded on the column and washed with another 50 mL of 0.5M HNO 3. This wasfollowed by the addition of 50 mL of water and 25 mL of 0.01M NH4OH. Themolybdenum was eluted with 25 mL of concentrated NH4OH. Again, 1-mL fractionsof wash and eluant were collected with the automatic sample collector. Thegamma activity and pH of each fraction were determined.
For these experiments, samples were counted both on theNaI detector and on a germanium detector equipped with an automatic samplechanger and a Nuclear Data 66 multichannel analyzer. For the analysis on the
117
NaI detector, the 9 9Mo doublet peak at 739 and 778 keV was used instead of the9 9Mo peak at 181 keV used previously to avoid interference with the 9 9Tc peakat 141 keV. The latter was the likely cause of (1) the >100% material balancefor 99 Mo and (2) the elution of the first peak attributed to 9 9Mo. Theresolution of the germanium detector allows separation of all of these peaksas singlets.
To determine the "neutralization point" of the column,another 3 g of alumina was prepared similarly. Then, the following solutionswere passed through the column: 100 mL of 0.5M HNO3 , 50 mL of water, and75 mL of 0.01M NH4OH. The ammonium solution was collected in 3-mL aliquotsand the pH of each was checked with pH paper. The first 22 samples had a pHof 5, while subsequent samples had a pH > 8. Therefore, the columnneutralized the first 66 mL of 0.01M NH4OH.
(2) Discussion
The ability of the alumina to neutralize the diluteammonium hydroxide explains the problem observed earlier of 9 9 Mo elution with0.01M NH4OH. An :.ncrease in pH does not signal the start of 0.01M NH4OHelution but rather:- the neutralization of the column. Therefore, the method ofusing pH measurements to determine when the 0.01M NH4OH was eluting isincorrect. The increase in pH indicates either the elution of concentratedNH4OH or the column being neutralized by the O.O1M NH4OH. In either case, therise in pH was accompanied by the elution of 9 9 Mo This suggests care must betaken not to wash with too much 0.01M NH4OH to avoid diluting
99 Moprematurely.
The graph of a run is presented in Fig. 11-30. The risein pH that accompanied the elution of 9 9Mo can be seen. It should be notedthat the pH was measured with pH paper to find the point of pH change, not todetermine the pH accurately. The higher energy 9 9 Mo peaks around 750 keV seemto be more reliable, yielding an elution of 96%, while the peak at 181 keVresulted in >100% elution. The full width at half maximum was 2.4 mL, while89% of the 99Mo was eluted in four column volumes. There is some tailing, buta 96% yield is reasonable.
3. Preparation and Irradiation of U3Si2
a. Target Preparation
Two miniplates of U3 Si2 have been identified as suitablefor production of 99Mo. They are 60-mil (1.5 mm) plates with 5.2 g/cm3
uranium. One miniplate was subsequently quartered, wrapped in aluminum foil,and electron-beam welded inside 1100-aluminum capsules. The aluminum capsuleswere then tested for He leakage. Due to the great amount of heat (6500 J, ora 175 C temperature rise for 27 g of aluminum) produced during an 18-minirradiation of 0.7 g of 2 3 5U in the ANL Janus reactor with a flux of2.35 x 1011 n/cm2 -s, each 1100-Al capsule was placed inside an aluminum rabittube for irradiation.
118
z0+s 12
-- cpm-PH
10
10+5 -- 8 =
6
000 -4
0 10 20 30 40 50 60
Volume of Col mn Effluent,mL
Fig. 11-30. Alumina Column Elution of 99 Mo. Runon 12/11/89. Counted on NaI detectorusing 9 9Mo doublet at 750 keV.
b. Experimental Irradiation
The uranium silicide sample was irradiated in the Janurreactor for 80 min at approximately 1/4 power (49.73 kW). As a check on thepower level, the inlet and outlet temperatures of the reactor water weremeasured. The temperature rise agrees with the power at which the reactor wasset.
The temperature of the sample was monitored using athermocouple. Initially, the air flow in the rabit tube for cooling was atits maximum, and the temperature went from 25*C to 29C over 40 min. At thistime, the air flow was decreased by switching to full bypass, and thetemperature went up to 31*C.
4. Chemical Processing, Separation, and Analysis of Irradiated U3 Si2
a. Chemical Processing
The day following the irradiation, the sample was received byus. After removal from the irradiation capsule and encasing aluminum foil,the sample's radioactivity was measured to be 1.3 R (gamma) and 18 R (gammaand beta) at 6 cm.
The sample was put into a 400 mL beaker with 20 mL of 3M NaOHand heated. The aluminum cladding began dissolving and a black precipitateformed. Another 43 mL of 3M NaOH was .dded, and the black precipitate andsolution were poured into two plastic centrifuge tubes. Then, 16 mL of3M NaOH was added to the sample and heated. Since reaction was observed, all
119
the aluminum was assumed to be dissolved. This solution was also added to thetwo tubes. The combined supernate in the tubes after centrifuging measured60 mL in a graduated cylinder. The cladding hydroxide precipitate waspartially dissolved in 1:1 HNO3 . Samples of both supernate and precipitatewere taken for gamma analysis.
On the third day, the remaining uranium silicide measured130 mR (gamma) and 180 mR (gamma and beta). The sample was dissolved byadding 20-mL aliquots of 1:1 3M NaOH/30% H202 and heating. After the reactionslowed down, the solution was decanted before adding fresh solvent. Thevolume of the final solution was 400 mL. Sodium nitrite crystals were addedto speed the destruction of the peroxide.
The original intent was to acidify the solution at this pointso that a homogeneous sample could be taken for counting. However, uponacidification with HNO 3, a white precipitate began forming. At the time, itwas believed to be the silica. Therefore, the solution was made basic byaddition of NaOH, and the hydroxides (including uranium) were centrifuged out.The total wet weight of this yellow precipitate was 63.17 g. This precipitatewas dissolved in 75 mL of 1:1 HNO3 . Brown fumes were observed at this point.The final solution was yellow and diluted to 100 mL in a volumetric flask.
The supernate from the uranium hydroxide precipitation wasacidified with HNO3 . Again, brown fumes were observed. At a pH of about 9, awhite fluffy precipitate began forming. As the solution approachedneutrality, there was considerable precipitate. Upon further acidification,the precipitate disappeared. A precipitate formed again at a pH of 6; andafter adding about 150 mL of 1:1 HNO3, the solution cleared, the pH dropped,and the solution became a light yellow. This molybdenum-containing solutionwas evaporated to a total volume of 500.00 mL at a pH of 0.310.
b. Separation of Molybdenum on Alumina Columns
Four grams of alumina was washed with water and put on a columnwith adjustable plungers at the top and bottom. The flow rate was set at1.2 mL/min. The column was not thermostated to avoid contamination ofequipment with the radioactive material. The alumina was pre-equilibrated bypassing 100 mL of 0.5M HN13 through the column.
About 200 mL of the molybdenum solution was loaded onto thecolumn, while the remaining 300 mL was further evaporated to 150 mL, thuslowering its pH. The fraction collector was used for the initial solutioncoming off the column as well as the later eluants. After the remaining300 mL was loaded on the alumina, the column was washed with 25 mL of 0.5MHNO3 , followed by 25 mL of H20. Potential impurities were eluted with 50 mLof 0.O1M NH4OH. The molybdate was eluted with 25 mL of concentrated NH40H.
c. Countieg
Gamma radiation data were collected using an ND66 system with agermanium detector. The spectra were analyzed for peaks and isotopeidentification using GAMANAL, a gamma spectrum analysis program.3 4
120
d. Results
(1) Initial Separation
Aliquots were taken from each of the four sample fractions(supernates and precipitates of both dissolved cladding and the uraniumsilicide fuel) for analysis by gamma spectroscopy. Final results for theseinitial separations are presented in Table 11-32.
With the exception of 239Np, all of the isotopes found arefission products. Apparently, activation of minor metals in the cladding isnot a major problem as far as radionuclidic purity.
Results show the presence of 99 Mo in all four fractions,with most of it in the supernates as expected. Since neither precipitate waswashed after centrifuging, it is not surprising that a fraction of the 9 9 Moremained in the precipitate fractions. An important result of this experimentis the 20% loss of Mo to the cladding dissolution solution. Other isotopeswere also lost to the cladding in various amounts. Xenon losses varied from66 to 100%. Because xenon is a gas and cannot be quantitatively accountedfor, this could be due to greater mobility within the sample or simply to thefuel samples being counted at a later date, after the xenon had escaped.About 20% of the iodine and cesium was lost to the cladding, indicatingrelative mobilities of these elements comparable to mol bdenum. Barium andthe lanthanides lost around 10%, except for 158 Eu and 1 3Sm, which lost 28 and2%, respectively. The gamma peaks for these isotopes are very close, andincorrect assignment of counts may explain these variations.
Recounting 42 days after the irradiation, when much of theshort-lived activity had decayed, allowed better analyses of the longer-livedisotopes, especially those of lower activity. This is particularly evidentfor 137Cs and 108Ru, which were not detected in the early countings. Theexpected activity of 2 3 9Np was calculated to be about 3 mCi six times higherthan that observed (possibly due to interference from the 1b2Te peak, which is0.2 keV away).
From the activity of each fission isotope, the moles offissioned 2 3 5 U can be calculated, correcting for fission yield. This numbershould theoretically be the same for all fission isotopes. The activities ofthe fission products indicated that about 3.2 x 1010 mol of 2 3 5 U fissioned.Initial analysis of 9 5Nb showed a lower number of 2 3 5 U moles fissioning, butlater counts are much closer to the average. Ruthenium-106 activity was toolow to be seen in the early countings and yielded an unexpected fissioning235U number about four times larger than average. The number of 2 3 5 U moles
fissioned as calculated from 1 33 Xe is smaller than average by about 200 timesbecause of xenon loss from the sample as it was being processed.
The total activity of 9 9 Mo produced (0.891 mCi) indicatesa flux of 5.1 x 1010, while the unperturbed flux was 6.5 x 1010 n/cm2-s. Thisis a flux depression of 78%.
(2) Alumina Column Separations
Analysis of the fuel supernate after acidification showsthat 79% of the iodine was lost during the acidification. In some processingsolvents, this iodine would be trapped and purified for later sale.
Table 11-32. Janus Irradiation: Initial Separations
dpaa
Dissolved Cladding U3Si2 FuelIrrad.,a aol U-235 % Lost to
Supernate Precip. Supernate ppt Total aCi Fissioned Cladding Half-life
Sr 91Nb 95b
Zr 95bZr 97Yo 99
Ru 103b
Rh 10bRu 106
bTc 131Tc 132I 131bI 133
Ic 133I 135
Ic 135Cs 137
bBa 140
bLa 140Ce 141
bCe 143Ce 144
bNd 147
bSa 153Eu 156Np 230
c
1.37E+95.58E+75.79E+74.68E+8
1.05E+6
2.32E+82.39E+82.30E+84.98E+92.14E+7
1.31E+79.21E+79.55E+79.23E+76.40E+92.13E+81.12E+71 . 14E+74.52E+8
3.22E+6
7.40E+82.39E+72.35E+7
3.07E+5
3.80E+53.94E+73.41E+8
1.71E+5
5.79E+66.16E+75.94E+71.37E+94.24E+73.80E+91.26E+9
9.69E+4
1.39E+9
6.071+64.83E+66.05E+64.65E+85.38E+78.66E+68.87E+65.91E+7
2.60E+71.299+8
1.61E+71.60E+8
4.02E+73.96E+78.96E+71.71E+71.601E+74.74E+8
1.96E+61.67B+71.96E+72.87E+75.59E+73.39E+7
0.6260.006
0.045
3.1100.8910.034
0.441
1.39E+91.31E+79.85E+71.00E+89.87E+76.90E+91.98E+97.57E+77.83E+79.79E+8
4.27E+62.60E+71.11E+93.25E+83.29E+86.601+96.38E+73.80E+91.26E+9
5.77E+50.2104.68E+86.011+91.57E+81.58E+85.58E+91.74E+71.881+71.931+81.82E+81.291+92.01E+81.25E+9
4.59E-113.45E-103.52E-103.46E-10
3.08E-103.59E-103.72E-10
1.42E-09
3.10E-103.15E-10
7.42E-12
3.70E-10
3.22E-10
3.13E-103.15E-10
2.84E-103.07E-102.81E-102.65E-10
577
2011126
1001219232366
100100
17981
11108
109
112
283
9.5 h35.0 d
64.0 d
16.9 h66.02 h39.4 d
35.4 h367 d
30 h78 h8.040 d
20.9 h5.25 d6.61 h9.10 h
30.17 y
12.79 d
40.3 h32.5 d
33.0 h284 d
11.0 d
46.8 h15 d
235 d
Total 17.879
aAt I=0, end of irradiation.
bCounted 42 days after irradiation.
cActivation product of U-238; all other isotopes are fission products.
4.80E+54.26E+8 4.66E+8
4.28E+85.92E+91.40E+81.42E+8
1.48E+9 3.631+91.74E+71.68E+71.76E+81.621+81.261+91.45E+81.22E+9
0.0120.4990.146
2.9280.0291.7120.568
3.22E-10
2.7070.071
2.5150.008
0.087
0.5800.0900.565
122
Fractions of the eluant from the aluminum columnseparation were taken for gamma analysis. Figure 11-31 shows that technetiumadsorption is minimal as it steadily elutes. From 350 to 375 mL, technetiumelution decreases during the 0.5M HNO3 acid wash. This is followed by a waterwash, during which technetium seems to begin eluting again. The Gaussian peakfrom 400 to 450 mL indicates technetium elution with the 0.01M NH4OH. Themolybdenum elutes abruptly upon the addition of concentrated NH4OH, and itshigher peaks could not be shown to scale in this plot. The elution ofmolybdenum was not followed long enough to observe the tailing of this peakbut, based on the observed counts, the molybdenum yield was >83%.
Both Ru aad Te (Fig. 11-32) pass through the aluminacolumn and appear to be completely washed out before the elution ofmolybdenum, thus ensuring separation of these isotopes from the 99Mo product.Iodine seems to be more of a problem. Although it passes through the column(Fig. 11-33), apparently some of the iodine does adsorb to elute with theconcentrated NH40H.
4e+7
+ Tc/mL
3e+/ - - Mo/mL
+
1. 2+7 -
2+l1e7 + +
++ +
0 100 200 300 400 500
Volume of Column Effluent,mL
Fig. 11-31. Technetium and Molybdenum Behavior on Alumina Column
mL
0-350 : loading solution350-375 :0.5M HNO3 wash375-400 H20 wash400-450 : 0.01M NH4OH wash>450 : conc. NH4OH elution
123
3e+5
E.
V
2e+5
to+5
OeO4
-
-D-- Ru-103
-*--- T.-132
-L-AL
0 100 200 300 400 5
Volume of Column Effluent,mL
Fig. 11-32. Ruthenium and Tellurium Behavior on Al
2.00.+6
C)
1.000+6 F
0.009+00 100 200 300 400 500
Volume of Column Effluent,mL
Fig. 11-33. Iodine Behavior on Alumina Column
00
umina Column
1 I _ I
_
124
5. Test of Acid Dissolution for Unirradiated U3 Si2
Unirradiated uranium silicide was dissolved in 3M HNO 3, leavinga white, finely divided precipitate. Analysis by X-ray diffraction showedthat the precipitate is amorphous and is probably silica. Silicon metal canalso be amorphous, but this is not likely under the oxidizing conditionsencountered here to form metallic silicon. Problems reported by others3 1 maybe radiation related.
6. Future Work
More irradiations will be performed in the future, anddifferent means of dissolving the irradiated LEU silicide will be tested.
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12. F. J. Millero, Water and Aqueous Solutions: Structure, Thermodynamics,and Transport Processes, R. A. Horne, Ed., Wiley-Interscience, New York,p. 13 (1972).
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14. A. Roux, G. M. Musbally, G. Perron, J. E. Desnoyers, P. P. Singh,E. M. Woolley, and L. G. Hepler, "Apparent Molal Heat Capacities andVolumes of Aqueous Electrolytes at 25 C: NaC103 , NaC104, NaNO3 , NaBr03 ,NaI03, KC103 , KBrO3 , KI 3, NH4NO3, NH4C1, and NH4C1O 4," Can. J. Chem. 56,24 (1978).
15. 0. Sohnel and P. Novotny, Densities of Aqueous Solutions of InorganicSubstances, Elsevier, New York (1985).
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25. K. L. Nash, R. C. Gatrone, G. A. Clark, P. G. Rickert, and E. P. Horwitz,"Hydrolytic and Radiolytic Degradation of OOD(iB)CMPO: ContinuingStudies," Sep. Sci. Tech. 23(12&13), 1355-1372 (1988).
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33. K. A. Burrill and R. J. Harrison, "Development of the 9 9Mo Process atCRNL," Fission Molybdenum for Medical Use, International Atomic EnergyAgency Report IAEA-TECDOC-515 (June 1989).
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127
III. HIGH LEVEL WASTE/REPOSITORY INTERACTIONS(J. K. Bates)
The Nevada Nuclear Waste Storage Investigation (NNWSI) Project isinvestigating the tuff beds of Yucca Mountain, Nevada, as a potential locationfor a high-level radioactive waste repository. As part of the waste packagedevelopment portion of this project, which is directed by Lawrence LivermoreNational Laboratory (LLNL), experiments are being performed by CMT to studythe behavior of the waste form under anticipated repository conditions. Theseexperiments include (1) the development and performance of a test to measurewaste form behavior in unsaturated conditions, (2) the performance ofexperiments designed to study the behavior of waste package components in anirradiated environment, and (3) the performance of experiments to investigatethe reaction of glass with water.
A. NNWSI Unsaturated Test Method(J. K. Bates, T. J. Gerding)
Briefly, the Unsaturated Test consists of periodically drippingrepository water onto a waste package assembly consisting of a glass wasteform and perforated, sensitized 304L stainless steel. In batch tests, thetests are terminated and the components analyzed, while in continuous tests asmall amount of test solution is analyzed every 13 weeks and the testcontinues. At this point, the N2 continuous tests have been completed throughthe 117-week sampling period. All the batch tests have been completed, andthree continuous tests and one blank test are ongoing. A description of thetest method and results through 65 weeks has been compiled into a topicalreport (in press) and a symposium presentation.' The solution resultsdepicting the cumulative release of cations and transuranic elements areplotted in Figs. III-1 to 111-4. There are common trends observed in each ofthe three continuous tests, and these data will be of use in validating modelsof glass dissolution that are being developed by NNWSI.
B. NNWSI Parametric Experiments(J. K. Bates)
Because the NNWSI Unsaturated Test rigidly sets many of the testparameters, the effect that each parameter may have on the final radionucliderelease needs to be studied. This is being done in parametric experiments.Reported below are the results of hydrothermal and vapor experiments withATM-1c and WV 44 glasses and obsidian.
A set of experiments was performed on ATM-1c glass and obsidian to meas-ure the difference between the layer thickness formed through hydrothermalleaching (submersion of sample in deionized water) and that formed with vaporhydration (sample reacted with water vapor). It had been demonstrated pre-viously2 that, at equivalent temperatures, hydrothermal leaching with theseglasses results in a thicker reaction layer. This suggests that the rate-controlling mechanism is different under the two conditions. The presentexperiments are being conducted to test the vapor phase hydration method thatwill be used in an upcoming hydration test matrix.
The ATM-1c glass is known to react rapidly under hydrothermal conditions,but in previous hydration tests,3 similar PNL 76-68 type glasses have appearedresistant to vapor hydration. Obsidian is of interest because the hydrationrind dating method4 uses a hydrothermal treatment to determine hydration rate
128
N2-9
0M9
U
N2-988 88 108 128
TIME (ueeks)u
40-
30-
cc9 28-
00-
0
-1 84
-18.S 28 40 '6 88 '10s
0-c~ 8
-U -
-30-
e-4H .
e 20 40 69 '8
TIME (ueeks)
0Li
Ma
Si
188 128
TIME (ueeks)
Fig. III-1. Cation ReleaseNo. N2-9
from the N2 Test Series,
7 r
6-
4.
3-
2-
1-
10C
E4
a00
F-
-IB 20 40
M2-9
0B
Ca
120
0
72
-
II
, '9'
, '
129
12-10
28 48 68 88 188
0
U
M2-1128
4A
TIME (ueeks)
30-
V~ 28-c
En
- 18-0
r.
M2-10 -1
:
e-
82 I28
4 1 I I I I I1 2
48 68 88 188 128
TIME (ueeks)
0Li
t
Ma
Si
28 40 60 88 100 120
TIME (weeks)
Fig. 111-2. Cation Release from the N2 Test Series,No. N2-10
7~
61
5-
413-
2-
1-
*0
c
0r
/
e. ru~
-1 8
0B
Ca
0I-100-
-3ee-
0
-
I q
L.j
130
N2-12
2-v 1-
0
-1-
8 28 48 68 88 188 128
TIME (ueeks)
N2--12."
158-
188-
* 58-
cEs0
r. 8-
o-58-
,1 88-
-158-
-288.9 28 40 GO 80 100
0Mg
1tu
n2-12
2
0C
cq ii0Ev 10.
-1
e
e
8
0B
Ca
, , , , , , , , , , ,20 40 60 0 180 120
TIME (weeks)
0Li
Na
Si
-121 28
TIME (ueeks)
Fig. 111-3. Cation Release from the N2 Test Series,No. N2-12
, 8
-
-
T
131
Mepttun i um
48r
36-
26-
18-
- I8T 20
Release
48 68 88 100 12
TIME (ueeks)
40-
3v-i
20-
18-1
AMericium Releasee 28 49 68 88 188 12
48 6 0e0 a0e '
TIME (ueeks)
DM209
t0241
20
TIME (weeks)
Fig. 111-4. Transuranic Release from the N2 Test Series,Nos. N2-9, N2-10, N2-12
ODJ
1
W
0M209
t2010
12912
20.
1312111"
87076s432-
1-
xc
e'
e
-40
v,up
0
t29
N201
N12S12
Plutonium Release8
i
1
132
constants for natural glasses. However, if vapor hydration results in a dif-ferent rate constant, then the accuracy of age dating measurements would becalled into question. Also being tested is WV 44, which is similar to theATM-10 glass that will be used in the extended vapor phase experiments des-cribed in Sec. III.F, and some data on its rate of hydrothermal reaction arerequired.
The experimental matrix is given in Table III-1. The ATM-1c glass wassupplied by the Materials Characterization Center (MCC); this glass type hasbeen used previously in the gamma leach experimental matrix. The obsidian wasobtained frc.a Northwestern University and was analyzed using quantitativeenergy dispersive X-ray spectroscopy (EDS). The WV 44 glass was supplied byCatholic University. The composition of the obsidian and WV 44 glass is shownin Table 111-2. The composition of ATM-lc is reported in Ref. 5.
The reaction progress was measured by determining weight loss and thereaction layer thickness by use of scanning electron microscopy (SEM)/EDS andsecondary ion mass spectrometry (SIMS). The SEM/EDS was the primary techniquebecause it is anticipated that the reaction layers will be thick (>2 m) andwill be easily observed. The SIMS was used on an obsidian sample to determinewhether the hydration process results in any measurable change in elementalprofiles in obsidian.
1. Obsidian
Hydrothermal Leaching
The obsidian samples were leached for periods of 1 to ~11 days. Theextent of reaction was measured by weight loss, SEM/EDS examination, SIMS, andoptical microscopy. The weight loss results are shown in Table III-1. Theseresults indicate that some leaching or etching of the obsidian has occurred.The effects of such leaching are also evident in the SEM and optical micro-scope, where a white crust is evident on the circular edge of the glass afteronly 24.5 h and on the glass surface after 72 h of reaction. This white crustis what remains of the glass after considerable etching occurs. In the circu-lar rims, the etched pits are particularly evident, as shown in Fig. 111-5.The thickness of the etched regions was measured, and the mass loss and layerthickness are plotted in Figs. 111-6 and 111-7. A sample reacted for threedays was examined with SIMS. The profiles indicate a slight Si depletion inthe first 0.1 pm, with an equivalent Al enrichment in the same region. Sodiumshows a depletion for the first 0.3 m, after which the Na profile is constantwhile K shows an enrichment that levels off after -1.3 m of sputtering. Thesample was sputtered to a total depth of ~6 pm to be assured of sputteringthrough the hydrated layer in an effort to determine if a sharp interface withthe bulk glass would be observed. No sharp interface was observed for theelements profiled.
Samples 13 and 15 were thin sectioned and examined using transmittedlight, which is the standard method of detecting the hydration layer inobsidian samples. This method resulted in clearly visible birefringent layerswhen the samples were observed under crossed-polarized light. The layerthicknesses were ~2 and 4 pm for samples 13 and 15, respectively.
Table III-1.
Sample Glass Temp., Dur.,
No. Type C h
Experimental Matrix and Test Results for ATM-lc and Obsidian Experiments
VesselDate
In Time
Date
Out Time
SaMple MassgIn Out
A A
Weight, g___ Mass, pH Mass,
No. In Out g Out g x 18-5
Hydrothermal Experiment
12/18/87 12:00 0.23615 0.23026
12/19/87 11:30 0.20100 0.19462
12/21/87 12:10 0.21280 0.20647
12/23/87 0.20630 0.19661
12/28/87 9:00 0.23891 0.22679
12/18/87 12:00 0.16634 0.15547
12/19/87 11:30 0.16151 0.15909
12/21/87 12:10 0.16689 0.16464
12/23/87 0.15872 0.15588
12/28/87 9:00 0.13205 0.13222
1
2
3
4
6
6
7
8
9
10
1044.95
1048.20
1047.74
1065.72
1045.51
1045.26
1045.27
1045.38
1046.29
1045.42
1047.64
1055.54
1045.31
*
1046.25
1045.28
1046.18
1045.11
0
0
0
0
0
0
- 9.7 489
- 10.2 638
.10 633
.18 10.2 969
.20 9.1 1210
- 7.8 87
- 7.9 142
.10 225
.11 7.3 284
.31 9.0 473
Vapor Phase Experiment;
12/31/87 3:30 0.25187 0.2b_ '3 82
1/04/88 12:15 0.30408 0.30407 83
1/05/88 12:30 0.27213 ND s
1/07/88 12:30 0.28407 ND 85
1/11/88 12:30 0.24881 ND 86
12/31/87 3:30 0.14975 0.14967 82
1/04/88 12:15 0.15909 0.15914 83
1/05/88 12:30 0.16664 ND 84
1/07/88 12:30 6.16528 ND 85
1/11/88 12:31 0.16702 ND 86
aNV = Nevada Obsidian.
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
JIB
1
2
3
4
6
11
12
13
14
15
6
7
8
9
1i
16
17
18
19
20
ATM-ic
ATM-ic
ATM-1c
ATM-ic
ATM-ic
NVOa
NYO
NVO
NYD
NVO
ATM-ic
ATM-ic
ATM-1c
ATM-ic
ATM-ic
NVO
NVO
NVO
NVO
NVO
188
188
188
188
188
188
188
188
188
188
188
188
188
188
188
188
188
188
188
188
24.5
48
72.5
141
261.5
24.5
48
72.5
141
261.5
72
24
48
96
192
72
24
48
96
192
12/17/87
12/17/87
12/17/87
12/17/87
12/17/87
12/17/87
12/17/87
12/17/87
12/17/87
12/17/87
12/28/87
1/03/88
1/03/88
1/03/88
1/13/88
12/28/87
1/03/88
1/03/88
1/03/88
1/03/88
11:31
11:30
11:30
11:30
11:30
11:30
11:30
11:30
11:30
11:30
3:30
12:30
12:30
12:30
12:30
3:30
12:30
12:30
12:30
12:30
CA)CA)
134
Table III-2. Composition of Glasses Used in Parametric Experiments
Glass Oxide, wt %
Type Si Na Al Ca Fea K Mg Ti Li B P Othersb
WV-44 44.8 9.9 6.5 0.9 11.6 3.2 1.2 0.9 2.8 9.0 2.3 6.9
Nevada
Obsid. 77.0 3.4 13.6 0.7 1.0 3.7 0.4 0.2 - - -
aFe as Fe 2 0 3 .
bIncludes Ba, Ce, Cr, Cs, Mn, Nd, Ni, S, Th, U, and Zr.
Fig. III-5. Cross Section of an Obsidian Sample Reactedfor 48 Hours in Deionized Water at 187*C.The dark areas penetrating into the glassare where the glass has dissolved.Magnification = 200X.
135
Fig. 111-6.
Mass Loss during Hydrothermal LeachingExperiments for ATM-ic Glass (squares)and Obsidian (circles)
1200
2900
E I
90
6500~
300
0.
0
0
00
03
00
0
4 8 12TIME (hours) 1/2
0
0
0
Fig. III-7.
Layer Thickness for Hydrothermal Leach-ing Experiments for ATM-ic Glass(circles) and Obsidian (squares)
o
4 BTIME
01
12(hours) i/2
0
0
16
50 140-
LJO
z 30.
S20
10
0 0
0
016
-v
v
136
Vapor Hydration
The glass surfaces, when examined with optical microscopy and SEM,appear unreacted. However, SIMS profiles indicate there is definite depletionof Ca and Na in the near-surface region extending ~0.3 m into the glass. Theextent of Na and Ca depletion appeared to be the same for both samples 18 and19 (48- and 96-h reaction).
2. ATM-1c
Hydrothermal Leaching
Previous leaching experiments done at 900C and higher temperaturesindicated that ATM-ic glass reacts quite rapidly in a leaching environment.In the present experiments the extent of reaction was monitored by measuringweight loss and reacted layer thickness, as shown in Figs. III-6 and III-7.Both give some indication that the growth of the reaction layer follows t'/2kinetics throughout the test period, reaching a layer thickness of ~51 pmafter 261.5 h.
Vapor Hydration
The vapor-hydrated samples showed little visible evidence forreaction. The exceptions were those samples which were watermarked,indicating that during the reaction period some water had condensed andcollected near the bottom of the sample. These samples were examined in theoptical microscope, SEM, and SIMS. Optical microscopy indicated that therewas no formation of reaction products on the glass surface, as was evidentwith SRL or WV glass under similar conditions. Scanning electron microscopyverifies this to be the case, with only the typical cross-hatched structurebeing evident (Fig. III-8). Also evident in this figure are small white
Fig. III-8. SEM Micrograph of ATM-ic Glass Reactedfor 48 Hours at 1870C in 100% Vapor
137
specks (<0.5 pm dia) that appear to be condensed regions. The EDS spectra ofthese specks, while difficult to obtain due to the small volume of the region,are not radically different from spectra of the general reacted surface.Thus, both optical and SEM examinations indicate the ATM-ic glass to berelatively unreacted in the vapor environment.
The SIMS technique was used to profile individual element behavior.Profiles for the samples reacted for one and two days are shown in Figs. 111-9and III-10, while the estimated depth at which bulk levels are reached isgiven in Table 111-3.
Several interesting features are evident in these profiles, whichbear further examination in the upcoming vapor hydration matrix. Theseinclude:
1. Sodium is depleted in the near-surface region but is depletedto neither the depth or extent (absolute Na signal) as are thehydrothermally leached samples or vapor-reacted SRL glasses.
2. The depletion depth ranges from ~0.1 to 1 pm over a period of96 h (Table 111-3), while the equivalent depletion in theleached samples is -51 pm (Fig. 111-7). This disparity inreaction between the vapor-reacted and leached samples is fargreater in ATM-1c glass than it is in SRL 165 glass.
3. In samples 7 and 9, there is a clear enhancement of the 45 amupeak at the glass surface. This peak is, in part, att:-ibutableto SiOH.
Thus, while the extent of reaction has not appeared to increase in auniform manner, sodium diffusion to the glass surface is less than would beexpected from previous hydration of either basaltic or other types of nuclearwaste glasses. One possible explanation is that the ability of a certainelement to diffuse or ion exchange with H+ or H2O+ depends on the chemicalpotential of that element at the glass surface. If a particular element, asit is released from the glass, is incorporated into a stable alterationproduct or goes into solution, the chemical potential for that element at thesurface is lower than in the bulk glass and the element is released from thebulk. However, if during vapor phase hydration, an element is released to thesurface and collects in the surface film but is not incorporated into analteration phase, the chemical potential may exceed that in the glass, therebyinhibiting further release of that element from the glass.
Examination of the surfaces of reacted glasses revealed noalteration phases.
3. WV 44
Vapor Hydration
The WV 44 glass reacted to form a distinct altered layer penetratinginto the glass, with alteration products forming on its surface. The layerthickness increased with time, as shown in Fig. III-11. There was an initialperiod where no distinct layer was observed in the SEM and no alterationproducts formed on the glass. However, after four days at 2000C, the glass
138
DEPTH (pm)
.
U7
0F-
0
4r2.5
.1087.
051-OS.
5 10 15 20 23 30
SPECTRUM #
DEPTH (pm;10'
p c./si-
6.TS/ss.
0 Fe/St.
(Ii
(1)
0I-
0H-
F-Q
3 10 20
SPECTRUM #
DEPTH (pm)0
JKB7.
015 10 15 20 25 30
SPECTRUM #
DEPTH (pm)0
JKO7.
AMg/St.
A 510/s I.
5 0 15 20
SPECTRUM #25 30
DEPTH (pm)
5 10 15 20
SPECTRUM #
Fig. 111-9. SIMS Profile of ATM-lc Reacted in Saturated WaterVapor for 24 Hours at 187*C
250000 4
1 4
?00000
HU)z 150000
I
100000
s0000
U)
0
0f--
.!U
0F-
0HI-
JIK7.
0B/S1.
AK/SI.
25 30
,aA--*: "
u j " " " " " " " " " " " " " " " " "
- - -"-
0
i
25 30
139
DEPTH (pm) DEPTH (pm)
mDCu
I
U)
O
cc
p51-2e.
5 10 15 20 25
SPECTRUM #
DEPTH (pm)so I
F-
0F-
Hl
5 10 15
SPECTRUM #20 25
2.5.
5 10 15SPECTRUM #
DEPTH (pm)
.1o 1
0sM/SI.
05 10 15
SPECTRUM #
DEPTH (pm).07
10 15
SPECTRUM #
Fig. III-10. SIMS Profile of ATM-ic Glass Reacted at 187 C inSaturated Water for 48 Hours
50000
CI)zLii
F-25000ZH--
.c8/9.
QNa/st.
.7,
.6
.S5
aJCu
0
F--
0
20 25
Jee
0Ca/St.
0 Fe/SI.
.3
.f
.1
20 25
wDCu
H
0H-
0I
.06
.05
.04
.03
.02
.010e/SI.
AK/li.
20 25
V L " " " " " " " f " " f " " "- :--" " -- -" "--"--- f--f---"-
V L " " " " " " " " " " " f "
"'__
0 s0
U
140
Table 111-3. Elemental Depletion Depths for Vapor PhaseReacted ATM-1c Glass
Reaction JKB Depth to Reach Bulk Concentration pmTime, Sampleh No.a B Na Ca SiOHb
24 7 0.8 0.3 0.5 Considerable
48 8 0.5 0.2 0.4 None
72 6 1.0 0.3 0.1 A Little
96 9 >0.7 0.1 0.2 Considerable
aSee Table III-1 for test description.
bQualitative assessment.
1000
CLWOE
75.
50
25
0
0
0
0
Fig. III-11.
Layer Thickness of Vapor (circles) andHydrothermally Reacted (squares) WV 44Glass (200*C)
0 0"
I n
0 5TIME
10(days)
.
began to rlact more rapidly, with extensive phase formation occurring as thelayer became thicker. At least seven different phases formed, includinganalcime, apatite, and phases that incorporate both uranium and thorium.
Similar reaction trends have been observed for the early generationwaste glasses SRL 211 and 131. For these glasses, the more rapid reactionappears to correspond to the formation of tobermorite, [Ca5Si6 01 6 (OH) 2 4H 20],on the glass surfaces
Hydrothermal Leaching
A distinct reaction layer also forms in the hydrothermal experimentswith WV 44. Unlike the vapor experiments, no distinct secondary phases formon the reacted surface. The layer increases in thickness at a fairly uniformrate (Fig. III-11) and does not show the change in slope observed in the vaporexperiments. Depending on the length of reaction, the layer thicknessproduced in the vapor experiments may exceed that of layers formedhydrothermally.
C. Relative Humidity and Simple Glass Experiments(B. M. Biwer)
A 28-d leach test following MCC-1 protocol (900C, surface-area-to-volumeratio of 0.01 mm-1) was conducted. It was designed to examine the effect of
0 o
mr
141
nonbridging oxygen ions on the process of hydration with glasses similar tothose used for nuclear waste. The composition of the glasses (Table 111-4)was varied systematically to decrease the interconnectivity of the silicanetwork. The concentration of the network modifier CaO was increased from itsinitial value of 5 mol % (Expt. 1) to 30 mol % (Expt. 11), which provides fora large number of nonbridging oxygens.
Table 111-4. Glass Compositions for Simple Glass Leach Test
Oxide Mole Percent
Glass No. SiO2 B2O3 A12 03 Na2O CaO
1 50 18.75 6.25 20 5
2 50 16.87 5.63 20 7.5
3 50 15.0 5.0 20 10
5 50 11.25 3.75 20 15
7 50 7.5 2.5 20 20
9 50 3.75 1.25 20 25
11 50 - - 20 30
The SEM/EDS results on the leached glasses revealed at least two, if notthree, layers on the surface of glasses 1, 2, 3, and 5. Depths of reactionfor glasses 1, 2, 3, and 5 were found to be 4, 3, 3, and 4 ,am, respectively.Glasses 7 and 9 were not examined since their reacted layers were flaking offthe surface, and the exposed underlying glass layer did not seem to have avery thick transition to bulk compositional values (as observed with SIMS).Glass 11 had a ceramic appearance and was not homogeneous when viewed throughthe optical microscope.
As shown in Fig. 111-12, the EDS signal for Ca in glass 1 increased ongoing from the bulk glass to a maximum, then decreased until it dropped offcompletely at the sample edge. These two regions were also clearly seen inthe Si and Al profiles. The Si concentration was depleted in both regionswhile Al was enriched. Sodium and boron were almost totally depleted in theselayers as might be expected for a leached glass.7 The surprising result isthe enrichment of boron just below the outer two layers.
With glasses 2, 3, and 5, two outer surface layers were always observed.Silicon was always depleted with respect to the bulk in both layers, as wasNa. Aluminum enrichment in the two regions occurred again in glass 2, but byglasses 3 and 5, where the aluminum oxide molar concentration is 5%, Al wasdepleted instead of enriched. The Ca concentration was enriched for the innerlayer and then depleted in the outer layer for glasses 3 and 5 but not 2,where it had a concentration comparable to that in the bulk. Boron was seento be depleted in the two surface layers for all glasses.
We attempted SIMS analysis of the reacted layers on glasses 1, 2, and 3.However, the SIMS profiles through the entire reacted layer were not obtaineddue to depth of reaction. Sodium and boron were depleted at the surface inall three cases. Subsurface enrichment of Ca and Al was observed forglasses 1 and 2 with surface enrichment for both on glass 3.
142
61R1P6
12S00
COUNTS
288
COUNTS
MICROMETERS IE
C1R1P6
BulkGlass
Al
Ca
l~y1Mounting
p ppox
MICROMETERS 16
Fig. 111-12. EDS Line Profile for Glass No. 1 Cross Section
a
0
Sii
Bulk
Glass
Ca
B1 1 Moun t i ng
.AEpoxy
I
143
Glass 5 had a reaction layer ~4-pm thick and was not profiled using SIMS.Glasses 7 and 9 had reacted layers which began flaking off he surface afterdrying. The exposed glass underneath in each case was profiled; depletion ofB, Na, and Ca with a possible enrichment of Si was seen. X-ray diffraction(XRD) of both the reacted film and the underlying glass for glass 7 indicatedthat there are no crystalline phases present. Analysis of the reacted filmfor glass 7 with EDS showed the presence of Si and Ca.
By increasing the number of nonbridging oxygens in the glass network withthe addition of CaO, one might expect faster hydration of the glass because amore open network structure would allow water more freedom to penetrate intothe glass. On the other hand, it must be recognized that addition of limitedamounts of CaO to a simple sodium silicate glass actually decreases thehydration rate of the glass, 8-1 0 despite the larger number of nonbridgingoxygens present. The Ca ion essentially plugs holes in the glass network andsterically impedes ion/water diffusion into, and ion migration out of, theglass. Such an effect could explain why the observed reaction layer forglass 1 was thicker than that seen on glass 2 (4 vs. 3 m) with significantlyhigher solution concentrations despite a smaller Ca concentration in theglass. The addition of more CaO (glass 3) results in a reacted layerthickness similar to that on glass 2 (with higher solution concentrations).This demonstrates that any beneficial effects initially gained are beginningto become outweighed by the further breakup of the glass network.
The existence of the Ca-enriched inner layer observed on glasses 1, 2, 3,and 5 suggests that the element forms a barrier to diffusion and could slowthe reaction rate of hydration. It is not clear whether the Ca enrichment isdue to ion migration from the outer layer to compensate for the leached Naions or from a precipitated phase such as CaSiO3 , as suggested by Clarket al.1 1 Clark et al. observed a CaO-enriched layer in a Na-depleted zonebelow the surface of a 20%Na2 O10%CaO'70%Si02 (mol %) glass leached for 9 daysat 100 C with a SA/V ratio of 0.77 cm-1 . The solution concentrations for Siand Ca in our experiments are not high enough for precipitation of Si,O,Hphases or Ca,O,H phases. A mixed phase such as CaSiO3 is possible.Experiments with a lower SA/V ratio would eliminate the possibility ofprecipitation and also the Ca-enriched layer if it is a precipitate phase.Unfortunately, interactions with Al and B cannot be ruled out.
The EDS line profiles provide evidence for buildup of B below the outertwo layers on the surface, where it is completely depleted, in what might becalled a third layer or region. This buildup could help explain the lower orcomparable B-to-Si ratios in solution vs. the bulk values (Table 111-5)despite the absence of B in the outer two layers. Why B segregates in such amanner is not clear at this time. It is known that some borosilicates undergophase separation under certain leaching conditions,7 and a B-enriched phasecould be forming at the glass/reacted layer interface. However, Bunker didnot report such phenomenon in his leaching study of Na 2 B2 0'SiO2 glasses. Itwas noted that the hydrolysis of B-O-B bonds occurs more readily than Si-O-Sibonds. The Si-O-Si bonds are, therefore, expected to be more resistant tohydrolysis than Si-O-B bonds. The addition of B203 to a sodium silicate, atthe expense of Si02 , then produces a glass more reactive toward hydrolysis.
Addition of A12 03 to a sodium silicate has the opposite effect andproduces a glass less reactive to hydrolysis.12 Enrichment of Al in the twosurface layers for glasses 1 and 2 is not surprising due to the low
144
Table 111-5. Elemental Mole Ratio to Silicon in Solution
Test
Vessels B Al Na Ca
#1 Bulk 0.75 0.25 0.80 0.10AG-1A 0.63 2 x 10-3 1.13 0.07AG-1B 0.61 4 x 10-3 1.02 0.08
#2 Bulk 0.67 0.23 0.80 0.15AG-2A 0.64 7.6 x 10-3 1.26 0.11AG-2B 0.61 8.2 x 10-3 1.10 0.12
#3 Bulk 0.60 0.20 0.80 0.20AG-3A 0.59 - 1.09 0.15AG-3B 0.56 4 x 10-3 1.20 0.17
#5 Bulk 0.45 0.15 0.80 0.30AG-5A 0.38 - 0.99 0.23AG-5B 0.35 - 0.89 0.24
#7 Bulk 0.30 0.10 0.80 0.40AG-7A 0.33 - 1.27 0.27AG-7B 0.30 - 1.06 0.32
#9 Bulk 0.15 0.05 0.80 0.50AG-9A 0.21 - 1.53 0.26AG-9B 0.21 - 1.36 0.25
#11 Bulk - - 0.80 0.60AG-11A - - 0.57 0.21
solubilities of Al oxides and hydroxides in aqueous solution. It is notenriched at the surface for glasses 3 and 5, but there is also less insolution.
Four points must be taken into consideration when analyzing the abovedata on going from glass 1 to glass 11, where CaO is substituted for A12 03 andB203 . First, Ca is a network modifier and can make the network more reactivetoward hydrolysis (more nonbridging oxygens). Second, less Al is present,which causes the glass to be more reactive toward hydrolysis. The last twopoints cover the opposite effect. Third, Ca can also decrease reactivitytoward hydrolysis (plugs pores). Fourth, the decreasing B content decreasesthe reactivity toward hydrolysis. Unfortunately, these effects are for sodiumsilicate glasses with the addition of Al, B, or Ca. All possible interactionsin the five-component glasses (SiO2 'B2 O3'Al2 O3'Na2O'CaO) are not known and canchange the effect that one component may have on the glass as a whole. Forexample, the beneficial effects of A1203 and CaO are not additive when bothare present in a sodium silicate.13 More work is needed on four-componentglasses before any attempts can be made at predicting (understanding) theleaching characteristics of a five-component glass such as this. However, thepresent samples will serve as references for further studies using vaporconditions and for examination of chemical bonding using spectroscopictechniques.
145
D. Analytical Support(W. L. Ebert and B. M. Biwer)
1. Development of SIMS Analysis Instrumentation
Secondary ion mass spectroscopy (SIMS) is being used to profileelements in reacted glass surfaces. The profiles are used to qualitativelydescribe the elemental distributions in the near-surface region and in thealteration layers which form during glass reaction. These concentrationprofiles can be used to measure the thickness of the alteration layer(s) ifthe sputter rate is known. Although SIMS results have been presented inprevious reports as well as this report, work continues to improve on both theacquisition and interpretation of the data. The SIMS technique has proven tobe very helpful in understanding the glass reaction.
2. Development of a Laser Raman Microprobe System
A laser Raman microprobe (LRM) system is being developed to analyzethe microcrystalline precipitates which form during the hydration of nuclearwaste glass. By using standard microscope objectives, an argon ion laser canbe focused to a beam diameter of less than 2 m. This allows spectra of verysmall samples to be collected with little or no background interference.X-ray emission spectroscopy, availale on the SEM, can be used to composition-ally analyze a sample. The Raman 'pectra of standards having compositionsconsistent with that of an unknown sample can then be compared to the Ramanspectrum of the unknown for comparative identification. The apparatus wascalibrated and tuned to improve the accuracy and resolution of the spectra.Computer collection of spectral data is being developed to allow spectralsmoothing, averaging, and background subtraction. The wavenumber readout ofthe monochromator used in the Raman microprobe apparatus has been calibratedto better than 1 cm-1 in the region of interest, from Ar+ excitation at19435.1 cm-1 (514.532 nm)14 through 15435.1 cm- 1 (A 4000 cm-1). The neon lampof an electrical line voltage tester was used as the calibration source. Boththe entrance and exit slits were set to 50 m (spectral bandpass 0 0.5 cm-1).
A polynomial regression analysis to the second power in the observedvalue was used to fit the data. The program POLYFIT was written in BASIC todo this with an Apple Ilgs computer. The actual value in absolute wavenumbers(Y) of an observed value (X) can be calculated according to the followingequation:
Y = (3.41573 x 10-7 )X2 + (0.964268)X + 528.0 (III-1)
The presence of a concave mirror on the laser grating filter causeddivergence of the laser beam. The entrance slit of the microscope acted as anaperture which restricted the amount of light incident on the sample. As aresult, all Raman work has been done with a laser power of 1 mW at thesample. The concave mirror was replaced by a plane mirror on the lasergrating filter. The plane mirror maintains the collimation of the laser andallows the entire laser beam to enter the microprobe. The laser power at thesample is now 10-20 mW with this modification.
The digital output of the photon counter has been interfaced to theApple IIgs computer. The program RAMAN for acquiring Raman spectra on thecomputer is being developed. Currently, it is running under the ProDOS8
146
operating system and will collect spectral data, store them, and then retrievethem from computer disks. Because of memory limitations of the currentoperating system, hardcopy and data manipulation options are not yetavailable. A new BASIC software package which uses the available memory moreefficiently has been received, and an updated program is under development.
E. Basalt Analog(J. J. Mazer)
A series of hydrothermal leaching and vapor phase hydration experimentswas performed by Byers et al.15 using two synthetic basalts and one SRL glassand deionized water. The composition of the three glasses was given in theprevious semiannual. While these experiments were completed under a differentsponsor, they provide a wealth of samples and data that still require analysisand interpretation, where the synthesized information will be of use to theNNWSI program. For this reason work is continuing with these samples.
Two test matrices were completed by Byers et al.15; in one matrix theglasses were leached at temperatures ranging from 90C to 1870C for times upto 546 days, and in the second the glasses were vapor hydrated at temperaturesranging from 1220 to 2400C for up to 150 days. The solutions were analyzedfor the cation concentrations, and the surfaces of the reacted glasses werecharacterized to identify mineral phases that had formed. The reacted glassesare presently being quantitatively analyzed with respect to the layercomposition and growth kinetics of the layer.
To determine the reaction kinetics of both the hydrothermal and vaporhydration experiments, the layer thickness on each of the reacted glasses wasmeasured. The layer thickness on individual samples at a given time variedsomewhat, more so in the vapor hydration experiments than in the leach tests.Results for Hawaiian basalt reacted at 1870 are presented in Fig. 111-13.Because of the scatter in the vapor hydration tests, these experiments arebeing repeated. It was noted in the original experiments that a significantamount of water had condensed on the glass. This is being avoided in therepeat experiments by preheating the samples before adding water. Maintainingwater saturation requires at least 0.16 g H2 0 in the vessel. In several casesthis condition was not maintained and the layer thicknesses varied greatly.In the cases where vapor saturation was continuous, the variations decreasedsignificantly. Figure 111-14 summarizes the vapor hydration experimentsperformed to date for Hawaiian basalt at 2000C. With the experimentaltechnique sufficiently refined, these experiments are being repeated. Thelayers will then be quantitatively analyzed in order to determine the reactedlayer composition as a function of time and the growth rate. This will alsofurther our understanding of how well leach tests and vapor tests can serve toaccelerate reactions.
F. Vapor Hydration Experiments(W. L. Ebert)
Knowledge of the secondary phases which form when nuclear waste glassesreact wita water is very important to modelers attempting to predict the ex-tent of reaction many hundreds of years into the future. It is necessary to
147
Hawaiian Basalt Reacted Layer Thicknesses
Q Leach
A Vapor Hydration
AA
AA
Ii15 2010
(T IME, d ay s)
Fig. 111-13. Layer Thicknesses as a Function of Time for HawaiianBasalt Reacted at 1870C. Open symbols denote thick-nesses from leach experiments, while filled symbolsrepresent thicknesses from vapor hydration experiments.The error bars represent the range of layer thicknesseson single sample.
7.0 m(0.147)
Hawaiian Basalt
Vapor Hydrated
200 C (0.196) (0.196)
(0.143)}-(0150)
I(0.142)I(0.190)
- (0.179)-1 -
10 20TIME, days
30
Fig. 111-14. Layer Thickness vs. Time for Hawaiian Basalt VaporHydrated at 200*C. Bars represent range of thicknessesmeasured. Number near each bar is the amount of 1120in grams reenId, in each vessel upon tterii nation of the
experiment.
10-
E
LLJ
64
4-2-
00
5-
LI-
3-
2-
1-
E
ofLUL
00
- -" D I
1T
lj
I
148
know, for example, which if any of the transuranic species may becomeincorporated into secondary minerals that precipitate from solution. Commonleaching studies used to measure glass dissolution rates are usually performedwith relatively low ratios for glass surface area to leachant volume. Thisprevents saturation of most precipitates during the duration of theexperiment. We have started a series of vapor hydration experiments whichavoid these problems.
In these experiments, glass samples are suspended in closed reactionbombs containing a small amount of water not in contact with the glass. Asthe reaction vessel is heated, the water vapor pressure within the bombincreases to the equilibrium vapor pressure, which is in excess of 14 atm at200"C. Experiments are performed at elevated temperatures in order toaccelerate the reactions. Under the high relative humidity in the vessel,water sorbs onto the glass surface to form a layer thought to be many tens ofmonolayers thick. The glass reaction can only proceed with this limitedamount of water, which provides a SA/V ratio in excess of several million percentimeter. This film becomes saturated with respect to several mineralsafter only a few days, the identity of the minerals being dependent on thecomposition of the glass.
Both hydrothermally reacted samples (wherein the sample is submerged inliquid leachate) and vapor reacted samples show an altered region on thesample surface. This is due to compositional changes that occur duringleaching and other glass reactions. The thickness of this region is oftenused as a measure of the extent of reaction. Figure 111-15 shows the measuredlayer thicknesses of hydrothermally and vapor reacted glass. The glass used
300-E
200-
H 100-
04c
/
5TIME,
10DAYS
Fig. 111-15.
Measured Alteration Layer Thickness asFunction of Reaction Time for WV 50Glass Reacted at 2000C with 0.20 g ofWater Added (squares) or Submerged inDeionized Water (circles). Symbolsshow the representative layer thicknessof individual samples, while the error
- bars denote the range of measured15 thicknesses on each sample.
in these experiments was WV 50 glass, a borosilicate glass representative ofthe West Valley Demonstration Plant feed. The hydrothermally reacted samplesfall on a straight line; all these samples have layers much thinner than thevapor reacted samples. The vapor reacted samples, for the most part, fall ona line with a much steeper slope. (A few experiments failed to produce areaction for unknown reasons. These results are not shown, but fall near thex-axis.) Other results imply that the extent of reaction in the vaporhydration experiments is dependent on the amount of water added to the vessel(see Sec. III.E). The low value at ten days of reaction or longer is probablydue to slight water loss during the experiment.
- - -
JLI
149
The layer thicknesses for the hydrothermally reacted samples may beunderestimated because of possible etching at the layer surface, but this isexpected to Le a small correction. It is speculated that the vapor hydratedsamples are jo much more reacted due to the ability of the pH to rise to morecorrosive values.
In the initial reaction, ion exchange occurs between a proton from waterand an alkali metal ion in the glass. The hydroxide produced by this reactioncan then catalyze hydrolysis of network bonds. A hydrothermal system tends todilute the hydroxide buildup because of the large volume of water available.Vapor reacted samples are unable to dilute the hydroxide buildup and so reachcorrosive pH values rapidly. As the glass reacts in the vapor environment,glass species are released into the aqueous film sorbed onto the surface,which soon becomes saturated with respect to different phases. These precip-itate onto the glass surface and a constant solution composition is attained.Although the data are shown to fall on a straight line in Fig. 111-15, it islikely that more data will tend to support t1" kinetics, with the rate-determining step being water infiltration into the glass. Further experimentsto determine the influence of the volume of water and the reaction kinetics ina saturated vapor environment are in progress.
G. Spent Fuel Leach Tests(E. Veleckis and J. C. Hoh)
A study is underway to determine radionuclide release rates from spentreactor fuels immersed in site-specific groundwater at ambient hot celltemperature. The study is a part of the NNWSI Project, whose responsibilityis to evaluate the performance of encased spent fuel as a high-level wasteform for permanent disposal in a potential repository located in the volcanictuff beds near Yucca Mountain, NV. Spent Fuel Leaching (SFL) tests aredesigned to simulate a possible condition in which groundwater may collect onthe bottom of a breached waste form canister during the post-containmentperiod. The study is a continuation of previous saturated SFL tests that havebeen carried out at Hanford Engineering and Development Laboratory (HEDL) in afour-series project. The current tests constitute Series 5 of the project.
Because of delays in procurement of J-13 well water, the activitiesrelated to SFL tests have been curtailed. In the interim period, we initiateda Reactant Dissolution Rate (RDR) experiment (QA level III) that is designedto test the feasibility of measuring the dissolution rate of a U0 2 matrix inwater by using isotope dilution techniques. According to the SFL test plan,similar measurements will be incorporated in one of the tests with spent fuel.The experimental setup and sampling procedures parallel those of SFL tests.The reactant consists of 80.2 g of uranium oxide powder with 15 wt % enrich-ment in 2 3 5 U and is contained in a bailed basket that has been enclosed in a304L stainless steel vessel filled with 250 mL of J-13 water. The experimentwill be carried out in two cycles. In the first cycle, the reactant U0 2 isleached in pure J-13 water until a steady-state concentration of uranium isestablished. In the second cycle, the leachant will be replaced with freshJ-13 water that has been spiked with depleted uranium at the steady-stateconcentration determined during the first cycle. The spiked solution will beassayed for total uranium, 235U content, and 2 3 8U/2 3 5U ratio.
150
The dissolution rate of the reactant, dN(t)/dt, will be calculated froman equation derived by Bruton and Shaw'6 :
dR1 (t)
dt(III-2)
where R1 (t) = Cia(t)/Clb(t) and Rr = Cra/Crb are the concentration ratiosbetween isotopes a and b for the solution phase (time dependent) and solidreactant phase (time independent), respectively; Ci(t) and Cri are theconcentrations of the i'th isotope in the solution (mol/L) and in the solidreactant (mol/g U02), respectively; N(t) is the weight of solid reactant(in andV is the solution volume (in L). Letters a and b stand for 2 3 8Uand U isotopes, respectively. Quantities Crb, V, and Rr have fixed values,whereas R1 (t) and Cib(t) will be determined as functions of time from theanalytical results. The dissolution rate of U02 can then be determined fromthe slope of the linear portion of the dR1 (t)/dt vs. (R1 (t) - Rr)/Cib(t) plotaccording to Eq. 111-2.
Dissolution results obtained during the first 63 days are shown inFig. 111-16. The samples are divided into two categories: those that arefiltered (through Amicon CF25 membrane cones or 0.4 m filter) and those thatare unfiltered. Sparging, filtration through the 0.4 m filter, and conesoaking appear to have negligible effect on the uranium content within theprecision of the data. The unfiltered samples contain particulate matter that
Fig. 111-16.
Uranium Concentration Data Availableto Date for the Leachate Samples ofthe RDR Experiment
250
0
x200 -
0
E
0
c
U
E0'c
150
100
Sporged, unfiltered
Quiescent, unfilteredSparged, 0.4 micron filterSparged, CF25 filter (not presooked)Sporged, CF25 filter (presooked)
~0
- o
_ o
- 0
I I I I I I I I I
50
n0 20 40 60
Time/days80 100
. .I I ; i i I i
v
ICr dN(t) R (t) - Rr
I i i I dtC (t)
151
becomes visible when settled in unused, unacidified leachate residues. Theestimated porosity of CF25 cone filter is 1.8 nm, and it is assumed that thesefiltrates contain the dissolved species only.
Uranium concentrations shown in Fig. 111-16 for CF25 filter fractions aremuch higher than those reported in the HEDL work. A plausible explanation ofthis observation is provided by the higher dissolution rates expected from thelarge surface area of the finely divided reactant used in the presentexperiment. According to a recent simulation study using the EQ3/6geochemical model on the dissolution of 100 g of spent fuel in 1 kg of J-13water at 25 0C,' 6 three U-bearing minerals were found to play a role incontrolling the uranium concentration in the leachate: haiweeite (38 ppb U),soddyite (60 ppb U), and schoepite (95 ppm U). The geochemical model predictsthat these three secondary phases appear (in the sequence indicated) duringthe dissolution of spent fuel and that emergence of the schoepite (U03'2H20)causes a 1000-fold increase in the uranium concentration. Because of therapid dissolution rate of U02 powder in the present experiment, the reactantmay have reached the composition where the uranium concentration is beingincreasingly controlled by the schoepite, thereby yielding values similar tothose predicted in the simulation study. Additional uranium concentration dataare being collected to verify the validity of this interpretation before thesecond cycle of the RDR experiment is initiated.
H. Radiation Effects(D. T. Reed)
The NNWSI project is investigating the feasibility of locating a nuclearhigh-level waste repository in the tuff formations in southern Nevada. Theplacement of high-level waste containers in the underground facility willperturb the pre-emplacement environment by raising the ambient temperature andexposing the environment to gamma radiation levels that initially may be inexcess of 0.1 x 106 rad/h. Determining the effects of temperature and gammaradiation on the gaseous/aqueous environment of the waste package is importantin characterizing the performance of the high-level waste package during thecontainment period in the repository history.
In FY 1988, the emphasis of the radiation effects task at CMT is toinvestigate the radiation chemistry of moist air systems. The work plannedconsists of performing a baseline series of experiments and investigatingthree related issues. The baseline set of experiments will determine the NOyield as a function of temperature and absorbed dose in 304L stainless steelvessels under conditions expected in a tuff environment. The three relatedissues are (1) the effect of (vessel surface area)/(irradiated gas volume)ratio on NO. yield, (2) the effect of water vapor content on NO yield, and(3) the effect of the candidate copper-based materials (copper nickel, oxygen-free copper, aluminum bronze) on the NO. yield.
In this report period, a series of 14 experiments in 304L stainless steelvessels and two experiments in Cu/Ni vessels was completed. These experimentswere performed at 30 and 1500C for both dry and water saturated (at roomtemperature) air. The objective of these experiments was to (1) develop andevaluate the test procedure and (2) provide preliminary indications on thechemistry of the irradiated system. The end product was the establishment ofa general rocedure for experiments on gas phase radiolysis.
152
The first test matrix completed according to the general procedure wasthe irradiation of dry air at room temperature (28 2'C) in 304L stainlesssteel vessels. The combined results of the post-irradiation analyses (gaschromatography and mass spectrometry) are given in Table 111-6.
The experiments performed covered the range of absorbed doses from 4.3 to308 Mrad at dose rates in the range of 0.1 to 0.4 Mrad/h and were 1 day up to3 months in duration. Seven absorbed doses were done in replicate to helpidentify extraneous results. Error due to temperature variation should beminimal since these were room-temperature experiments. Some variability mayhave been introduced due to unavoidable delays in sample analysis with themass spectrometer following sample concentration.
Radiolytically, these experiments represent a measurement of initialyield data, with the exception that the results have been obtained overrelatively long times. Typical initial yield studies published are on theorder of an hour or less in duration. Pulse radiolysis studies are usually onthe order of seconds or less. The data being generated in this work,therefore, included an unavoidable contribution from the vessel wall via bothcatalysis of intermediate reactions on the vessel surface and the removal ofproducts from the gas phase via sorption. This is viewed as an integralpart of the NNWSI problem, since NO. formation under repository-relevantconditions will occur over long times and in the proximity of potentiallyseveral surfaces (e.g., host rock, container liner, container material).
The yield of nitrous oxide was found to increase linearly with anincrease in absorbed dose (linearity coefficient, 0.993; slope,0.00257 mol %/Mrad). The yield of nitrous oxide, based on the slope obtained,was 0.56. This is significantly higher than the yield of 0.38-0.44 reportedelsewhere171"8 and is attributed to contributions from the reaction vesselwall.
More detailed study of the results indicated that there was an inversecorrelation between the yields of carbon dioxide and NO (NO + N20 + NO2 ).Unusually low NOX yields in runs SS-35, -34, and -38 were accompanied by thefcrmation of a relatively high amount of 002. Small amounts of 002 as acontaminant in the system are repository relevant but not expected tointerfere with the NO. chemistry of the system. The results, however,indicate that the precursor to 002, most likely an organic compound, had asignificant effect on the radiation chemistry observed. (Significant CO2formation was usually accompanied by significant lower N2 0 formation.)
As was observed in the procedure-test experiments, the yield of NO2 wasmore variable than that of N20. The measured gaseous concentrations in thereplicate experiments were within experimental error except when a substantialamount of 002 was generated. The effect of 002 generation on NO2 yield wassimilar to that observed in the case of N20. The NO2 yield was comparable toor greater than N20 yield, although a more quantitative comparison cannot bemade until the NO2 on the vessel walls has been included.
Nitric oxide yields, as observed in the procedure-test experiments, werenot correlated with absorbed dose. This is consistent with the characteriza-tion of nitric oxide as an intermediate which does not build up over time.
Table 111-6. Product Analyses of Dry Air Irradiation in 304L Vessels at 28 C
Product Analysis, mol %
N 2 02 CO2 Ar NO1
308 78.9 19.2 0.19 0.93 1.9 0.1
301 79.0 19.0 0.29 0.93 2.0 0.1
163 78.8 20.0 0.14 0.93 0.94 0.05
169 78.9 19.8 0.28 0.93 1.1 0.05
77.0 21.8 0.0653 0.93
77.3 21.5 0.23 0.93
76.9 22.0 0.028 60.9377.2 21.7 0.228 60.93
0.65 0.05
0.11 0.001
0.26 0.020.041 0.004
77.8 21.4 0.0650 0.93 0.042 0.004
77.4 21.8 60.029 6.93 0.082 0.009
77.6 21.5 60.015 60.93 0.019 6.004
77.5 21.6 0.012 0.93 0.017 6.003
20
23
21
22
33
35
32
34
38
39
41
40
36
37
0.81
0.74
1.1 0.05
1.2 0.08
0.39 0.52 0.03
0.31 0.74 0.04
0.17 0.48 0.040.036 0.068 0.005
0.085 0.18 0.013
0.015 60.022 0.002
0.010 0.032 0.003
0.020 0.0680 0.005
0.007 0.013 0.003
0.006 0.008 0.001
!8.088 !5.082!8.8834 !.003
<0.003
56.004
ND
ND
0.04 0.02
0.01 0.002
ND
0.005 0.001
ND
0.004 0.0008
ND
<0.003
ND
0.003 0.001
<0.001
<0.00007
0.01 0.007
0.01 0.007
0.02 0.006
0.02 0.005
0.02 0.006
0.01 0.002
0.005 0.002
0.003 0.001
0.005 0.002
0.005 0.002
0.002 0.0007
0.001 0.006
<0.003
<_.002
Run
No.
ss-
Absd.
Dose,
Mrad N2 0 NO2 NO
4.3 77.5 21.5 60.620 .93
4.3 77.5 21.5 60.15 0.93
H2 0
77
75
46
42
20
19
10
10
C)
154
Positive identification of ammonia as a radiolytic product was not made.This was not conclusive since the method of analysis was not very sensitive toammonia. Its formation is, however, not expected in a stainless steel air-system.
I. Spent Fuel Literature Review(D. T. Reed)
A review of the spent fuel literature is being done as part of the wasteform interaction subtask in the Materials Literature Review and Waste PackageStrategy Development Program supported by the Repository Technology Program.The objective of this subtask is to identify, collect, and evaluate existingliterature on the performance and characteristics of spent fuel under condi-tions relevant to a high-level nuclear waste repository. The emphasis hasbeen on the chemical durability/stability of spent fuel, its interaction withthe components of the waste package, and the related radiochemistry.
The data pertaining to spent fuel behavior in the natural environmenthave been almost exclusively generated within the nuclear high-level wasteprograms in the United States and Europe. There is not a broad base ofinformation that is directly relevant to spent fuel behavior in geologicmedia. There are, therefore, relatively few studies (on the order of 200 orless) in existence, and in many cases, the work performed has not undergonethe peer review and test of reproducibility that are the usual criteria forscientific validity. For this reason, all studies and reports that discussspent fuel are being evaluated without pre-screening them for their directrelevance to the interests of the Repository Technology Program.
Approximately 150 reports and articles that directly pertain to spentfuel performance under repository-relevant conditions have been identified. Asubstantial number of these are readily available and have been collected. Ofthese, most have been surveyed (informally reviewed), and over a hundred arebeing formally evaluated.
Efforts to identify and collect literature on issues that are indirectlyrelated to spent fuel performance (such as radiochemistry in groundwatersystems, self-irradiation effects, kinetics of uranium oxide dissolution,uranium ore body movement) have not been completed. These subtopics are notnecessarily applicable to all candidate repositories and evaluating theirimportance is part of the process of developing a strategy for spent fuelcharacterization work.
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